Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

DOE PAGES

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; ...

2017-01-24

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

Solid State Division progress report for period ending March 31, 1992

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

Green, P.H.; Hinton, L.W.

1992-09-01

During this period, the division conducted a broad, interdisciplinary materials research program with emphasis on theoretical solid state physics, superconductivity, neutron scattering, synthesis and characterization of materials, ion beam and laser processing, and the structure of solids and surfaces. The High Flux Isotope Reactor was returned to full operation.

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

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.

We report that many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has beenmore » examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Lastly, although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.« less

ZPR-6 assembly 7 high {sup 240} PU core : a cylindrical assemby with mixed (PU, U)-oxide fuel and a central high {sup 240} PU zone.

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

Lell, R. M.; Schaefer, R. W.; McKnight, R. D.

Over a period of 30 years more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited to form the basis for criticality safety benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactormore » physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. The term 'benchmark' in a ZPR program connotes a particularly simple loading aimed at gaining basic reactor physics insight, as opposed to studying a reactor design. In fact, the ZPR-6/7 Benchmark Assembly (Reference 1) had a very simple core unit cell assembled from plates of depleted uranium, sodium, iron oxide, U3O8, and plutonium. The ZPR-6/7 core cell-average composition is typical of the interior region of liquid-metal fast breeder reactors (LMFBRs) of the era. It was one part of the Demonstration Reactor Benchmark Program,a which provided integral experiments characterizing the important features of demonstration-size LMFBRs. As a benchmark, ZPR-6/7 was devoid of many 'real' reactor features, such as simulated control rods and multiple enrichment zones, in its reference form. Those kinds of features were investigated experimentally in variants of the reference ZPR-6/7 or in other critical assemblies in the Demonstration Reactor Benchmark Program.« less

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

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

Not Available

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for physical protection of irradiated reactor... Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1... of irradiated reactor fuel in excess of 100 grams in net weight of irradiated fuel, exclusive of...

Source Term Experiments Project (STEP): Aerosol characterization system

NASA Astrophysics Data System (ADS)

Schlenger, B. J.; Dunn, P. F.

A series of four experiments is being conducted at Argonne National Laboratory's TREAT Reactor. They were designed to provide some of the necessary data regarding magnitude and release rates of fission products from degraded fuel pins, physical and chemical characteristics of released fission products, and aerosol formation and transport phenomena. These are in pile experiments, whereby the test fuel is heated by neutron induced fission and subsequent clad oxidation in steam environments that simulate as closely as practical predicted reactor accident conditions. The test sequences cover a range of pressure and fuel heatup rate, and include the effect of Aq/In/Cd control rod material.

Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

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

Jennifer Lyons; Wade R. Marcum; Mark D. DeHart

2014-01-01

The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by themore » Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.« less

Five Lectures on Nuclear Reactors Presented at Cal Tech

DOE R&D Accomplishments Database

Weinberg, Alvin M.

1956-02-10

The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SPSE REACTOR ...

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

OVERVIEW OF NUCLEAR PHYSICS LABORATORY (IMMEDIATELY EAST OF SP-SE REACTOR ROOM), LEVEL -15’, LOOKING SOUTHWEST. NOTE SLIDING STEEL PLATE DOOR BETWEEN LABORATORY AND REACTOR ROOM - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

Scalable Methods for Uncertainty Quantification, Data Assimilation and Target Accuracy Assessment for Multi-Physics Advanced Simulation of Light Water Reactors

NASA Astrophysics Data System (ADS)

Khuwaileh, Bassam

High fidelity simulation of nuclear reactors entails large scale applications characterized with high dimensionality and tremendous complexity where various physics models are integrated in the form of coupled models (e.g. neutronic with thermal-hydraulic feedback). Each of the coupled modules represents a high fidelity formulation of the first principles governing the physics of interest. Therefore, new developments in high fidelity multi-physics simulation and the corresponding sensitivity/uncertainty quantification analysis are paramount to the development and competitiveness of reactors achieved through enhanced understanding of the design and safety margins. Accordingly, this dissertation introduces efficient and scalable algorithms for performing efficient Uncertainty Quantification (UQ), Data Assimilation (DA) and Target Accuracy Assessment (TAA) for large scale, multi-physics reactor design and safety problems. This dissertation builds upon previous efforts for adaptive core simulation and reduced order modeling algorithms and extends these efforts towards coupled multi-physics models with feedback. The core idea is to recast the reactor physics analysis in terms of reduced order models. This can be achieved via identifying the important/influential degrees of freedom (DoF) via the subspace analysis, such that the required analysis can be recast by considering the important DoF only. In this dissertation, efficient algorithms for lower dimensional subspace construction have been developed for single physics and multi-physics applications with feedback. Then the reduced subspace is used to solve realistic, large scale forward (UQ) and inverse problems (DA and TAA). Once the elite set of DoF is determined, the uncertainty/sensitivity/target accuracy assessment and data assimilation analysis can be performed accurately and efficiently for large scale, high dimensional multi-physics nuclear engineering applications. Hence, in this work a Karhunen-Loeve (KL) based algorithm previously developed to quantify the uncertainty for single physics models is extended for large scale multi-physics coupled problems with feedback effect. Moreover, a non-linear surrogate based UQ approach is developed, used and compared to performance of the KL approach and brute force Monte Carlo (MC) approach. On the other hand, an efficient Data Assimilation (DA) algorithm is developed to assess information about model's parameters: nuclear data cross-sections and thermal-hydraulics parameters. Two improvements are introduced in order to perform DA on the high dimensional problems. First, a goal-oriented surrogate model can be used to replace the original models in the depletion sequence (MPACT -- COBRA-TF - ORIGEN). Second, approximating the complex and high dimensional solution space with a lower dimensional subspace makes the sampling process necessary for DA possible for high dimensional problems. Moreover, safety analysis and design optimization depend on the accurate prediction of various reactor attributes. Predictions can be enhanced by reducing the uncertainty associated with the attributes of interest. Accordingly, an inverse problem can be defined and solved to assess the contributions from sources of uncertainty; and experimental effort can be subsequently directed to further improve the uncertainty associated with these sources. In this dissertation a subspace-based gradient-free and nonlinear algorithm for inverse uncertainty quantification namely the Target Accuracy Assessment (TAA) has been developed and tested. The ideas proposed in this dissertation were first validated using lattice physics applications simulated using SCALE6.1 package (Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) lattice models). Ultimately, the algorithms proposed her were applied to perform UQ and DA for assembly level (CASL progression problem number 6) and core wide problems representing Watts Bar Nuclear 1 (WBN1) for cycle 1 of depletion (CASL Progression Problem Number 9) modeled via simulated using VERA-CS which consists of several multi-physics coupled models. The analysis and algorithms developed in this dissertation were encoded and implemented in a newly developed tool kit algorithms for Reduced Order Modeling based Uncertainty/Sensitivity Estimator (ROMUSE).

An experimental test plan for the characterization of molten salt thermochemical properties in heat transport systems

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

Pattrick Calderoni

2010-09-01

Molten salts are considered within the Very High Temperature Reactor program as heat transfer media because of their intrinsically favorable thermo-physical properties at temperatures starting from 300 C and extending up to 1200 C. In this context two main applications of molten salt are considered, both involving fluoride-based materials: as primary coolants for a heterogeneous fuel reactor core and as secondary heat transport medium to a helium power cycle for electricity generation or other processing plants, such as hydrogen production. The reference design concept here considered is the Advanced High Temperature Reactor (AHTR), which is a large passively safe reactormore » that uses solid graphite-matrix coated-particle fuel (similar to that used in gas-cooled reactors) and a molten salt primary and secondary coolant with peak temperatures between 700 and 1000 C, depending upon the application. However, the considerations included in this report apply to any high temperature system employing fluoride salts as heat transfer fluid, including intermediate heat exchangers for gas-cooled reactor concepts and homogenous molten salt concepts, and extending also to fast reactors, accelerator-driven systems and fusion energy systems. The purpose of this report is to identify the technical issues related to the thermo-physical and thermo-chemical properties of the molten salts that would require experimental characterization in order to proceed with a credible design of heat transfer systems and their subsequent safety evaluation and licensing. In particular, the report outlines an experimental R&D test plan that would have to be incorporated as part of the design and operation of an engineering scaled facility aimed at validating molten salt heat transfer components, such as Intermediate Heat Exchangers. This report builds on a previous review of thermo-physical properties and thermo-chemical characteristics of candidate molten salt coolants that was generated as part of the same project [1]. However, this work focuses on two materials: the LiF-BeF2 eutectic (67 and 33 mol%, respectively, also known as flibe) as primary coolant and the LiF-NaF-KF eutectic (46.5, 11.5, and 52 mol%, respectively, also known as flinak) as secondary heat transport fluid. At first common issues are identified, involving the preparation and purification of the materials as well as the development of suitable diagnostics. Than issues specific to each material and its application are considered, with focus on the compatibility with structural materials and the extension of the existing properties database.« less

EBR-II Reactor Physics Benchmark Evaluation Report

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

Pope, Chad L.; Lum, Edward S; Stewart, Ryan

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

New Reactor Physics Benchmark Data in the March 2012 Edition of the IRPhEP Handbook

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

John D. Bess; J. Blair Briggs; Jim Gulliford

2012-11-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) was established to preserve integral reactor physics experimental data, including separate or special effects data for nuclear energy and technology applications. Numerous experiments that have been performed worldwide, represent a large investment of infrastructure, expertise, and cost, and are valuable resources of data for present and future research. These valuable assets provide the basis for recording, development, and validation of methods. If the experimental data are lost, the high cost to repeat many of these measurements may be prohibitive. The purpose of the IRPhEP is to provide an extensively peer-reviewed set ofmore » reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. Contributors from around the world collaborate in the evaluation and review of selected benchmark experiments for inclusion in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [1]. Several new evaluations have been prepared for inclusion in the March 2012 edition of the IRPhEP Handbook.« 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.

Radiation damage characterization in reactor pressure vessel steels with nonlinear ultrasound

NASA Astrophysics Data System (ADS)

Matlack, K. H.; Kim, J.-Y.; Wall, J. J.; Qu, J.; Jacobs, L. J.

2014-02-01

Nuclear generation currently accounts for roughly 20% of the US baseload power generation. Yet, many US nuclear plants are entering their first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. This means that critical components, such as the reactor pressure vessel (RPV), will be exposed to higher levels of radiation than they were originally intended to withstand. Radiation damage in reactor pressure vessel steels causes microstructural changes such as vacancy clusters, precipitates, dislocations, and interstitial loops that leave the material in an embrittled state. The development of a nondestructive evaluation technique to characterize the effect of radiation exposure on the properties of the RPV would allow estimation of the remaining integrity of the RPV with time. Recent research has shown that nonlinear ultrasound is sensitive to radiation damage. The physical effect monitored by nonlinear ultrasonic techniques is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features such as dislocations, precipitates, and their combinations. Current findings relating the measured acoustic nonlinearity parameter to increasing levels of neutron fluence for different representative RPV materials are presented.

REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

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

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

1970-05-01

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

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

Rim, Jung H.; Kuhn, Kevin J.; Tandon, Lav

Nuclear forensics techniques, including micro-XRF, gamma spectrometry, trace elemental analysis and isotopic/chronometric characterization were used to interrogate two, potentially related plutonium metal foils. These samples were submitted for analysis with only limited production information, and a comprehensive suite of forensic analyses were performed. Resulting analytical data was paired with available reactor model and historical information to provide insight into the materials’ properties, origins, and likely intended uses. Both were super-grade plutonium, containing less than 3% 240Pu, and age-dating suggested that most recent chemical purification occurred in 1948 and 1955 for the respective metals. Additional consideration of reactor modelling feedback andmore » trace elemental observables indicate plausible U.S. reactor origin associated with the Hanford site production efforts. In conclusion, based on this investigation, the most likely intended use for these plutonium foils was 239Pu fission foil targets for physics experiments, such as cross-section measurements, etc.« less

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.

REACTOR PHYSICS CONSTANTS

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

None

1963-07-01

This second edition is based on data available on March 15, 1961. Sections on constants necessary for the interpretation of experimental data and on digital computer programs for reactor design and reactor physics have been added. 1344 references. (D.C.W.)

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

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

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…

Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

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

Burkes, Douglas; Casella, Andrew M.; Buck, Edgar C.

2015-03-15

The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 onmore » four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents thermal property measurements conducted in FY 2014 and compares results to values obtained from literature and measurements performed in FY 2013, where applicable, along with appropriate discussion.« less

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.

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

10 CFR Appendix D to Part 73 - Physical Protection of Irradiated Reactor Fuel in Transit, Training Program Subject Schedule

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Physical Protection of Irradiated Reactor Fuel in Transit... Irradiated Reactor Fuel in Transit, Training Program Subject Schedule Pursuant to the provision of § 73.37 of... reactor fuel is required to assure that individuals used as shipment escorts have completed a training...

The Ongoing Impact of the U.S. Fast Reactor Integral Experiments Program

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

John D. Bess; Michael A. Pope; Harold F. McFarlane

2012-11-01

The creation of a large database of integral fast reactor physics experiments advanced nuclear science and technology in ways that were unachievable by less capital intensive and operationally challenging approaches. They enabled the compilation of integral physics benchmark data, validated (or not) analytical methods, and provided assurance of future rector designs The integral experiments performed at Argonne National Laboratory (ANL) represent decades of research performed to support fast reactor design and our understanding of neutronics behavior and reactor physics measurements. Experiments began in 1955 with the Zero Power Reactor No. 3 (ZPR-3) and terminated with the Zero Power Physics Reactormore » (ZPPR, originally the Zero Power Plutonium Reactor) in 1990 at the former ANL-West site in Idaho, which is now part of the Idaho National Laboratory (INL). Two additional critical assemblies, ZPR-6 and ZPR-9, operated at the ANL-East site in Illinois. A total of 128 fast reactor assemblies were constructed with these facilities [1]. The infrastructure and measurement capabilities are too expensive to be replicated in the modern era, making the integral database invaluable as the world pushes ahead with development of liquid metal cooled reactors.« less

The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes

NASA Astrophysics Data System (ADS)

Bogdanova, E. V.; Kuznetsov, A. N.

2017-01-01

The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.

75 FR 67636 - Physical Protection of Shipments of Irradiated Reactor Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-03

...-2010-0340; Draft NUREG-0561, Revision 2] RIN 3150-AI64 Physical Protection of Shipments of Irradiated...- 0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance to a licensee or applicant for implementation of proposed 10 CFR 73.37, ``Requirements for Physical...

78 FR 31821 - Physical Protection of Shipments of Irradiated Reactor Fuel

Federal Register 2010, 2011, 2012, 2013, 2014

2013-05-28

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 73 [NRC-2010-0340; NRC-2009-0163] RIN 3150-AI64 Physical..., ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This revised document sets forth means... physical protection of spent nuclear fuel (SNF) during transportation by road, rail, and water; and for...

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

Eugene P. Wigner's Visionary Contributions to Generations-I through IV Fission Reactors

NASA Astrophysics Data System (ADS)

Carré, Frank

2014-09-01

Among Europe's greatest scientists who fled to Britain and America in the 1930s, Eugene P. Wigner made instrumental advances in reactor physics, reactor design and technology, and spent nuclear fuel processing for both purposes of developing atomic weapons during world-war II and nuclear power afterwards. Wigner who had training in chemical engineering and self-education in physics first gained recognition for his remarkable articles and books on applications of Group theory to Quantum mechanics, Solid state physics and other topics that opened new branches of Physics.

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

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

Bess, J. D.; Briggs, J. B.; Gulliford, J.

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energymore » Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning is the critical experiments with fast reactor fuel rods in water, interesting in terms of justification of nuclear safety during transportation and storage of fresh and spent fuel. These reports provide a detailed review of the experiment, designate the area of their application and include results of calculations on modern systems of constants in comparison with the estimated experimental data.« less

High Throughput Plasma Water Treatment

NASA Astrophysics Data System (ADS)

Mujovic, Selman; Foster, John

2016-10-01

The troublesome emergence of new classes of micro-pollutants, such as pharmaceuticals and endocrine disruptors, poses challenges for conventional water treatment systems. In an effort to address these contaminants and to support water reuse in drought stricken regions, new technologies must be introduced. The interaction of water with plasma rapidly mineralizes organics by inducing advanced oxidation in addition to other chemical, physical and radiative processes. The primary barrier to the implementation of plasma-based water treatment is process volume scale up. In this work, we investigate a potentially scalable, high throughput plasma water reactor that utilizes a packed bed dielectric barrier-like geometry to maximize the plasma-water interface. Here, the water serves as the dielectric medium. High-speed imaging and emission spectroscopy are used to characterize the reactor discharges. Changes in methylene blue concentration and basic water parameters are mapped as a function of plasma treatment time. Experimental results are compared to electrostatic and plasma chemistry computations, which will provide insight into the reactor's operation so that efficiency can be assessed. Supported by NSF (CBET 1336375).

Determination of origin and intended use of plutonium metal using nuclear forensic techniques.

PubMed

Rim, Jung H; Kuhn, Kevin J; Tandon, Lav; Xu, Ning; Porterfield, Donivan R; Worley, Christopher G; Thomas, Mariam R; Spencer, Khalil J; Stanley, Floyd E; Lujan, Elmer J; Garduno, Katherine; Trellue, Holly R

2017-04-01

Nuclear forensics techniques, including micro-XRF, gamma spectrometry, trace elemental analysis and isotopic/chronometric characterization were used to interrogate two, potentially related plutonium metal foils. These samples were submitted for analysis with only limited production information, and a comprehensive suite of forensic analyses were performed. Resulting analytical data was paired with available reactor model and historical information to provide insight into the materials' properties, origins, and likely intended uses. Both were super-grade plutonium, containing less than 3% 240 Pu, and age-dating suggested that most recent chemical purification occurred in 1948 and 1955 for the respective metals. Additional consideration of reactor modeling feedback and trace elemental observables indicate plausible U.S. reactor origin associated with the Hanford site production efforts. Based on this investigation, the most likely intended use for these plutonium foils was 239 Pu fission foil targets for physics experiments, such as cross-section measurements, etc. Copyright © 2017 Elsevier B.V. All rights reserved.

Determination of origin and intended use of plutonium metal using nuclear forensic techniques

DOE PAGES

Rim, Jung H.; Kuhn, Kevin J.; Tandon, Lav; ...

2017-04-01

Nuclear forensics techniques, including micro-XRF, gamma spectrometry, trace elemental analysis and isotopic/chronometric characterization were used to interrogate two, potentially related plutonium metal foils. These samples were submitted for analysis with only limited production information, and a comprehensive suite of forensic analyses were performed. Resulting analytical data was paired with available reactor model and historical information to provide insight into the materials’ properties, origins, and likely intended uses. Both were super-grade plutonium, containing less than 3% 240Pu, and age-dating suggested that most recent chemical purification occurred in 1948 and 1955 for the respective metals. Additional consideration of reactor modelling feedback andmore » trace elemental observables indicate plausible U.S. reactor origin associated with the Hanford site production efforts. In conclusion, based on this investigation, the most likely intended use for these plutonium foils was 239Pu fission foil targets for physics experiments, such as cross-section measurements, etc.« less

Neutrino Physics with Nuclear Reactors: An Overview

NASA Astrophysics Data System (ADS)

Ochoa-Ricoux, J. P.

Nuclear reactors provide an excellent environment for studying neutrinos and continue to play a critical role in unveiling the secrets of these elusive particles. A rich experimental program with reactor antineutrinos is currently ongoing, and leads the way in precision measurements of several oscillation parameters and in searching for new physics, such as the existence of light sterile neutrinos. Ongoing experiments have also been able to measure the flux and spectral shape of reactor antineutrinos with unprecedented statistics and as a function of core fuel evolution, uncovering anomalies that will lead to new physics and/or to an improved understanding of antineutrino emission from nuclear reactors. The future looks bright, with an aggressive program of next generation reactor neutrino experiments that will go after some of the biggest open questions in the field. This includes the JUNO experiment, the largest liquid scintillator detector ever constructed which will push the limits of this detection technology.

75 FR 36126 - Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-24

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0228] Office of New Reactors; Proposed Revision to Standard Review Plan Section 13.6.1, Revision 1 on Physical Security--Combined License and Operating...), Section 13.6.1 on ``Physical Security--Combined License and Operating Reactors,'' (Agencywide Documents...

Sub 2 nm Particle Characterization in Systems with Aerosol Formation and Growth

NASA Astrophysics Data System (ADS)

Wang, Yang

Aerosol science and technology enable continual advances in material synthesis and atmospheric pollutant control. Among these advances, one important frontier is characterizing the initial stages of particle formation by real time measurement of particles below 2 nm in size. Sub 2 nm particles play important roles by acting as seeds for particle growth, ultimately determining the final properties of the generated particles. Tailoring nanoparticle properties requires a thorough understanding and precise control of the particle formation processes, which in turn requires characterizing nanoparticle formation from the initial stages. The knowledge on particle formation in early stages can also be applied in quantum dot synthesis and material doping. This dissertation pursued two approaches in investigating incipient particle characterization in systems with aerosol formation and growth: (1) using a high-resolution differential mobility analyzer (DMA) to measure the size distributions of sub 2 nm particles generated from high-temperature aerosol reactors, and (2) analyzing the physical and chemical pathways of aerosol formation during combustion. Part. 1. Particle size distributions reveal important information about particle formation dynamics. DMAs are widely utilized to measure particle size distributions. However, our knowledge of the initial stages of particle formation is incomplete, due to the Brownian broadening effects in conventional DMAs. The first part of this dissertation studied the applicability of high-resolution DMAs in characterizing sub 2 nm particles generated from high-temperature aerosol reactors, including a flame aerosol reactor (FLAR) and a furnace aerosol reactor (FUAR). Comparison against a conventional DMA (Nano DMA, Model 3085, TSI Inc.) demonstrated that the increased sheath flow rates and shortened residence time indeed greatly suppressed the diffusion broadening effect in a high-resolution DMA (half mini type). The incipient particle size distributions were discrete, suggesting the formation of stable clusters that may be intermediate phases between initial chemical reactions and downstream particle growth. The evolution of incipient cluster size distributions further provided information on the gaseous precursor reaction kinetics, which matched well with the data obtained through other techniques. Part 2. The size distributions and their evolution measured by the DMAs help explain the physical pathways of aerosol formation. The chemical analysis of the incipient particles is an important counterpart to the existing characterization method. The chemical compositions of charged species were measured online with an atmospheric pressure interface time-of-flight mass spectrometer (APi-TOF). The tandem arrangement of the high-resolution DMA and the APi-TOF realized the simultaneous measurement of the mobility and the mass of combustion-generated natively charged particles, which enabled their chemical and physical formation pathways to be derived. The results showed that the initial stages of particle formation were strongly influenced by chemically ionized species during combustion, and that incipient particles composed of pure oxides did not exist. The effective densities of the incipient particles were much lower than those of bulk materials, due to their amorphous structures and different chemical compositions. Measuring incipient particles with high-resolution DMAs is limited because a DMA classifies charged particles only, while the charging characteristics of sub 2 nm particles are not well understood. The charge fraction of combustion-generated incipient particles was measured by coupling a charged particle remover and a condensation particle counter. A high charge fraction was observed, confirming the strong interaction among chemically ionized species and formed particles. The combustion system was modeled by using a unimodal aerosol dynamics model combined with Fuchs' charging theory, and showed that the charging process indeed affected particle formation dynamics during combustion.

76 FR 5102 - Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated Reactor Fuel...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-28

... 3150-AI64 [NRC-2010-0340] Draft NUREG-0561, Revision 2; Physical Protection of Shipments of Irradiated...-0561, ``Physical Protection of Shipments of Irradiated Reactor Fuel.'' This document provides guidance on implementing the provisions of proposed 10 CFR Part 73.37, ``Requirements for Physical Protection...

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

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

John D. Bess; J. Blair Briggs; Jim Gulliford

2014-10-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

PubMed

Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-08-06

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

PubMed Central

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-01-01

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

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

Davis, R. S.

The following are specific topics of this paper: 1.There is much creativity in the manner in which Dimensional Generator can be applied to a specific programming task [2]. This paper tells how Dimensional Generator was applied to a reactor-physics task. 2. In this first practical use, Dimensional Generator itself proved not to need change, but a better user interface was found necessary, essentially because the relevance of Dimensional Generator to reactor physics was initially underestimated. It is briefly described. 3. The use of Dimensional Generator helps make reactor-physics source code somewhat simpler. That is explained here with brief examples frommore » BURFEL-PC and WIMSBURF. 4. Most importantly, with the help of Dimensional Generator, all erroneous physical expressions were automatically detected. The errors are detailed here (in spite of the author's embarrassment) because they show clearly, both in theory and in practice, how Dimensional Generator offers quality enhancement of reactor-physics programming. (authors)« less

Characterizing fluid dynamics in a bubble column aimed for the determination of reactive mass transfer

NASA Astrophysics Data System (ADS)

Kováts, Péter; Thévenin, Dominique; Zähringer, Katharina

2018-02-01

Bubble column reactors are multiphase reactors that are used in many process engineering applications. In these reactors a gas phase comes into contact with a fluid phase to initiate or support reactions. The transport process from the gas to the liquid phase is often the limiting factor. Characterizing this process is therefore essential for the optimization of multiphase reactors. For a better understanding of the transfer mechanisms and subsequent chemical reactions, a laboratory-scale bubble column reactor was investigated. First, to characterize the flow field in the reactor, two different methods have been applied. The shadowgraphy technique is used for the characterisation of the bubbles (bubble diameter, velocity, shape or position) for various process conditions. This technique is based on particle recognition with backlight illumination, combined with particle tracking velocimetry (PTV). The bubble trajectories in the column can also be obtained in this manner. Secondly, the liquid phase flow has been analysed by particle image velocimetry (PIV). The combination of both methods, delivering relevant information concerning disperse (bubbles) and continuous (liquid) phases, leads to a complete fluid dynamical characterization of the reactor, which is the pre-condition for the analysis of mass transfer between both phases.

RELAP-7 Development Updates

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

Zhang, Hongbin; Zhao, Haihua; Gleicher, Frederick Nathan

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory, and is the next generation tool in the RELAP reactor safety/systems analysis application series. RELAP-7 development began in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway of the Light Water Reactor Sustainability (LWRS) program. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. The code is beingmore » developed based on Idaho National Laboratory’s modern scientific software development framework – MOOSE (the Multi-Physics Object-Oriented Simulation Environment). The initial development goal of the RELAP-7 approach focused primarily on the development of an implicit algorithm capable of strong (nonlinear) coupling of the dependent hydrodynamic variables contained in the 1-D/2-D flow models with the various 0-D system reactor components that compose various boiling water reactor (BWR) and pressurized water reactor nuclear power plants (NPPs). During Fiscal Year (FY) 2015, the RELAP-7 code has been further improved with expanded capability to support boiling water reactor (BWR) and pressurized water reactor NPPs analysis. The accumulator model has been developed. The code has also been coupled with other MOOSE-based applications such as neutronics code RattleSnake and fuel performance code BISON to perform multiphysics analysis. A major design requirement for the implicit algorithm in RELAP-7 is that it is capable of second-order discretization accuracy in both space and time, which eliminates the traditional first-order approximation errors. The second-order temporal is achieved by a second-order backward temporal difference, and the one-dimensional second-order accurate spatial discretization is achieved with the Galerkin approximation of Lagrange finite elements. During FY-2015, we have done numerical verification work to verify that the RELAP-7 code indeed achieves 2nd-order accuracy in both time and space for single phase models at the system level.« less

Characterizations of the radioactive waste by the remotely-controlled collimated spectrometric system

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

Stepanov, Vyacheslav E.; Potapov, Victor N.; Smirnov, Sergey V.

Decontamination and decommissioning of the research reactors MR (Testing Reactor) and RFT (Reactor of Physics and Technology) has recently been initiated in the National Research Center (NRC) 'Kurchatov institute', Moscow. In the building, neighboring to the reactor, the storage of HLRW is located. The storage is made of monolithic concrete in which steel cells depth 4 m are located. In cells of storage the HLRW packed into cases are placed. These the radioactive waste are also subject to export on long storage in the specialized organization. For characterization of the radioactive waste in cases the remote-controlled collimated spectrometer system wasmore » used. The system consists of a spectrometric collimated gamma-ray detector, a color video camera and a control unit, mounted on a rotator, which are mounted on a tripod with the host computer. For determination of specific activity of radionuclides in cases, it is developed programs of calculation of coefficients of proportionality of specific activity to the corresponding speeds of the account in peaks of full absorption at single specific activity of radionuclides in cases. For determination of these coefficients the mathematical model of spectrometer system based on the Monte-Carlo method was used. Dependences of calibration coefficients for various radionuclides from distance between the detector and a case at various values of the radioactive waste density in cases are given. Measurements of specific activity in cases are taken and are discussed. By results of measurements decisions on the appeal of the radioactive waste being in cases are made. (authors)« less

Research and Engineering Operation, Irradiation Processing Department monthly record report, May 1965

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

Ambrose, T.W.

1965-06-04

Process and development activities reported include: depleted uranium irradiations, thoria irradiation, and hot die sizing. Reactor engineering activities include: brittle fracture of 190-C tanks, increased graphite temperature limits for the F reactor, VSR channel caulking, K reactor downcomer flow, zircaloy hydriding, and ribbed zircaloy process tubes. Reactor physics activities include: thoria irradiations, E-D irradiations, boiling protection with the high speed scanner, and in-core flux monitoring. Radiological engineering activities include: radiation control, classification, radiation occurrences, effluent activity data, and well car shielding. Process standards are listed, along with audits, and fuel failure experience. Operational physics and process physics studies are presented.more » Lastly, testing activities are detailed.« less

Solar physical vapor deposition preparation and microstructural characterization of TiO2 based nanophases for dye-sensitized solar cell applications.

PubMed

Negrea, Denis; Ducu, Catalin; Moga, Sorin; Malinovschi, Viorel; Monty, Claude J A; Vasile, Bogdan; Dorobantu, Dorel; Enachescu, Marian

2012-11-01

Titanium dioxide exists in three crystalline phases: anatase, rutile and brookite. Although rutile is thermodynamically more stable, anatase is considered as the most favorable phase for photocatalysis and solar energy conversion. Recent studies have shown a significant improvement of light harvesting and overall solar conversion efficiency of anatase nanoparticles in dye-sensitized solar cells (DSSCs) when using a mixture of anatase and rutile phases (10-15% rutile). TiO2 nanopowders have been prepared by a solar physical vapor deposition process (SPVD). This method has been developed in Odeillo-Font Romeu France using "heliotron" solar reactors working under concentrated sunlight in 2 kW solar furnaces. By controlling reactor's atmosphere type (air/argon) and gas pressure, several types of anatase/rutile nanophases have been obtained with slightly different microstructural properties and morphological characteristics. X-ray diffraction analyses (XRD) were performed on precursor and on the SPVD obtained nanopowders. Information concerning their phase composition and coherence diffraction domain (crystallites size and strain) was obtained. Nanopowders morphology has been studied by scanning electron microscopy (SEM) and transmission electron microscopy (TEM).

10 CFR 73.35 - Requirements for physical protection of irradiated reactor fuel (100 grams or less) in transit.

Code of Federal Regulations, 2014 CFR

2014-01-01

... fuel (100 grams or less) in transit. 73.35 Section 73.35 Energy NUCLEAR REGULATORY COMMISSION... Transit § 73.35 Requirements for physical protection of irradiated reactor fuel (100 grams or less) in... quantity of irradiated reactor fuel weighing 100 grams (0.22 pounds) or less in net weight of irradiated...

Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

DOE PAGES

Bess, John D.; Fujimoto, Nozomu

2014-10-09

Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in themore » experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

Detectability prediction for a thermoacoustic sensor in the breazeale nuclear reactor pool

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

Smith, James; Hrisko, Joshua; Garrett, Steven

2016-03-01

Laboratory experiments have suggested that thermoacoustic engines can be in- corporated within nuclear fuel rods. Such engines would radiate sounds that could be used to measure and acoustically-telemeter information about the op- eration of the nuclear reactor (e.g., coolant temperature or uxes of neutrons or other energetic particles) or the physical condition of the nuclear fuel itself (e.g., changes in temperature, evolved gases) that are encoded as the frequency and/or amplitude of the radiated sound [IEEE Measurement and Instrumen- tation 16(3), 18-25 (2013)]. For such acoustic information to be detectable, it is important to characterize the vibroacoustical environments within reactors.more » Measurements will be presented of the background noise spectra (with and with- out coolant pumps) and reverberation times within the 70,000 gallon pool that cools and shields the fuel in the 1 MW research reactor on Penn State's campus using two hydrophones, a piezoelectric projector, and an accelerometer. Sev- eral signal-processing techniques will be demonstrated to enhance the measured results. Background vibrational measurement were also taken at the 250 MW Advanced Test Reactor, located at the Idaho National Laboratory, using ac- celerometers mounted outside the reactor's pressure vessel and on plumbing will also be presented. The detectability predictions made in the thesis were validated in September 2015 using a nuclear ssion-heated thermoacoustic sensor that was placed in the core of the Breazeale Nuclear Reactor on Penn State's campus. Some features of the thermoacoustic device used in that experiment will also be revealed. [Work supported by the U.S. Department of Energy.]« less

Analytical Data Report for Sediment Samples Collected From 116-KE-3 and UPR-100-K-1; Boreholes C8796 and C8796

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

Snyder, Michelle M.V.; Last, George V.; Stephenson, John R.

2016-03-01

CH2M Hill Plateau Remediation Company (CHPRC) requested the services of the Pacific Northwest National Laboratory (PNNL) to perform contaminant leach testing on samples from two boreholes, C8796 and C8797, installed near the 105-KE reactor. These tests consisted of field texture column tests, <2 mm repacked column tests, batch desorption tests, and ion exchange experiments. In addition, hydraulic and physical property characterization was performed.

Developments in Sensitivity Methodologies and the Validation of Reactor Physics Calculations

DOE PAGES

Palmiotti, Giuseppe; Salvatores, Massimo

2012-01-01

The sensitivity methodologies have been a remarkable story when adopted in the reactor physics field. Sensitivity coefficients can be used for different objectives like uncertainty estimates, design optimization, determination of target accuracy requirements, adjustment of input parameters, and evaluations of the representativity of an experiment with respect to a reference design configuration. A review of the methods used is provided, and several examples illustrate the success of the methodology in reactor physics. A new application as the improvement of nuclear basic parameters using integral experiments is also described.

On fundamental quality of fission chain reaction to oppose rapid runaways of nuclear reactors

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, E. G.

2017-01-01

It has been shown that the in-hour equation characterizes the barriers and resistibility of fission chain reaction (FCR) against rapid runaways in nuclear reactors. Traditionally, nuclear reactors are characterized by the presence of barriers based on delayed and prompt neutrons. A new barrier based on the reflector neutrons that can occur when the fast reactor core is surrounded by a weakly absorbing neutron reflector with heavy atomic weight was proposed. It has been shown that the safety of this fast reactor is substantially improved, and considerable elongation of prompt neutron lifetime "devalues" the role of delayed neutron fraction as the maximum permissible reactivity for the reactor safety.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

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

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... nonpower reactors licensed to operate at or above a power level of 2 megawatts thermal. [38 FR 35430, Dec... OF PLANTS AND MATERIALS Physical Protection Requirements at Fixed Sites § 73.60 Additional...

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

Communicate the fact that high-power water-cooled test reactors such as the Advanced Test Reactor (ATR), the High Flux Isotope Reactor (HFIR) or the Jules Horowitz Reactor (JHR) cannot provide fast flux levels as high as sodium-cooled fast test reactors. The memo first presents some basics physics considerations about neutron fluxes in test reactors and then uses ATR, HFIR and JHR as an illustration of the performance of modern high-power water-cooled test reactors.

Specifications for a coupled neutronics thermal-hydraulics SFR test case

NASA Astrophysics Data System (ADS)

Tassone, A.; Smirnov, A. D.; Tikhomirov, G. V.

2017-01-01

Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors are a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed. Two sub-assemblies (SA) characterized by different fuel enrichment and layout are considered. Specifications for the test case are provided including geometrical data, material compositions, thermo-physical properties and coupling scheme details. Serpent and ANSYS-CFX are used as reference in the description of suitable inputs for the performing of the benchmark, but the use of other code combinations for the purpose of validation of the results is encouraged. The expected outcome of the test case are the axial distribution of volumetric power generation term (q‴), density and temperature for the fuel, the cladding and the coolant.

Risk Informed Margins Management as part of Risk Informed Safety Margin Characterization

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

Curtis Smith

2014-06-01

The ability to better characterize and quantify safety margin is important to improved decision making about Light Water Reactor (LWR) design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margin management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. In addition, as research and development in the LWR Sustainability (LWRS) Program and other collaborative efforts yield new data, sensors, and improved scientific understanding of physical processes that govern the aging and degradation of plant SSCs needs and opportunities to better optimize plantmore » safety and performance will become known. To support decision making related to economics, readability, and safety, the Risk Informed Safety Margin Characterization (RISMC) Pathway provides methods and tools that enable mitigation options known as risk informed margins management (RIMM) strategies.« less

Growth and Characterization of III-V Semiconductors for Device Applications

NASA Technical Reports Server (NTRS)

Williams, Michael D.

2000-01-01

The research goal was to achieve a fundamental understanding of the physical processes occurring at the surfaces and interfaces of epitaxially grown InGaAs/GaAs (100) heterostructures. This will facilitate the development of quantum well devices for infrared optical applications and provide quantitative descriptions of key phenomena which impact their performance. Devices impacted include high-speed laser diodes and modulators for fiber optic communications at 1.55 micron wavelengths and intersub-band lasers for longer infrared wavelengths. The phenomenon of interest studied was the migration of indium in InGaAs structures. This work centered on the molecular beam epitaxy reactor and characterization apparatus donated to CAU by AT&T Bell Laboratories. The material characterization tool employed was secondary ion mass spectrometry. The training of graduate and undergraduate students was an integral part of this program. The graduate students received a thorough exposure to state-of-the-art techniques and equipment for semiconductor materials analysis as part of the Master''s degree requirement in physics. The undergraduates were exposed to a minority scientist who has an excellent track record in this area. They also had the opportunity to explore surface physics as a career option. The results of the scientific work was published in a refereed journal and several talks were presented professional conferences and academic seminars.

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

NASA Technical Reports Server (NTRS)

Roth, R. J.

1976-01-01

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

Reactor Application for Coaching Newbies

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

2015-06-17

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

78 FR 69139 - Physical Security-Design Certification and Operating Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-18

... scheduled to close on October 30, 2013. The Nuclear Energy Institute (NEI) submitted a letter on October 9... NUCLEAR REGULATORY COMMISSION [NRC-2013-0225] Physical Security--Design Certification and Operating Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard review plan--draft section...

Breeder Reactors, Understanding the Atom Series.

ERIC Educational Resources Information Center

Mitchell, Walter, III; Turner, Stanley E.

The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

Thermal-hydraulic interfacing code modules for CANDU reactors

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

Liu, W.S.; Gold, M.; Sills, H.

1997-07-01

The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

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

Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.

2009-10-09

The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected tomore » come from increasingly diverse educational and experiential backgrounds.« less

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

This project details re-evaluations of experiments of gas-cooled fast reactor (GCFR) core designs performed in the 1970s at the PROTEUS reactor and create a series of International Reactor Physics Experiment Evaluation Project (IRPhEP) benchmarks. Currently there are no gas-cooled fast reactor (GCFR) experiments available in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). These experiments are excellent candidates for reanalysis and development of multiple benchmarks because these experiments provide high-quality integral nuclear data relevant to the validation and refinement of thorium, neptunium, uranium, plutonium, iron, and graphite cross sections. It would be cost prohibitive to reproduce suchmore » a comprehensive suite of experimental data to support any future GCFR endeavors.« less

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license...

Advances in boron neutron capture therapy (BNCT) at kyoto university - From reactor-based BNCT to accelerator-based BNCT

NASA Astrophysics Data System (ADS)

Sakurai, Yoshinori; Tanaka, Hiroki; Takata, Takushi; Fujimoto, Nozomi; Suzuki, Minoru; Masunaga, Shinichiro; Kinashi, Yuko; Kondo, Natsuko; Narabayashi, Masaru; Nakagawa, Yosuke; Watanabe, Tsubasa; Ono, Koji; Maruhashi, Akira

2015-07-01

At the Kyoto University Research Reactor Institute (KURRI), a clinical study of boron neutron capture therapy (BNCT) using a neutron irradiation facility installed at the research nuclear reactor has been regularly performed since February 1990. As of November 2014, 510 clinical irradiations were carried out using the reactor-based system. The world's first accelerator-based neutron irradiation system for BNCT clinical irradiation was completed at this institute in early 2009, and the clinical trial using this system was started in 2012. A shift of BCNT from special particle therapy to a general one is now in progress. To promote and support this shift, improvements to the irradiation system, as well as its preparation, and improvements in the physical engineering and the medical physics processes, such as dosimetry systems and quality assurance programs, must be considered. The recent advances in BNCT at KURRI are reported here with a focus on physical engineering and medical physics topics.

The Virtual Environment for Reactor Applications (VERA): Design and architecture

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

Turner, John A., E-mail: turnerja@ornl.gov; Clarno, Kevin; Sieger, Matt

VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goalsmore » and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.« less

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS

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

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Manymore » research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.« less

Characterization of cross-section correction to charge exchange recombination spectroscopy rotation measurements using co- and counter-neutral-beam views.

PubMed

Solomon, W M; Burrell, K H; Feder, R; Nagy, A; Gohil, P; Groebner, R J

2008-10-01

Measurements of rotation using charge exchange recombination spectroscopy can be affected by the energy dependence of the charge exchange cross section. On DIII-D, the associated correction to the rotation can exceed 100 kms at high temperatures. In reactor-relevant low rotation conditions, the correction can be several times larger than the actual plasma rotation and therefore must be carefully validated. New chords have been added to the DIII-D CER diagnostic to view the counter-neutral-beam line. The addition of these views allows determination of the toroidal rotation without depending on detailed atomic physics calculations, while also allowing experimental characterization of the atomic physics. A database of rotation comparisons from the two views shows that the calculated cross-section correction can adequately describe the measurements, although there is a tendency for "overcorrection." In cases where accuracy better than about 15% is desired, relying on calculation of the cross-section correction may be insufficient.

High Pressure Low Temperature X-Ray Diffraction Studies of UO2 and UN single crystals.

NASA Astrophysics Data System (ADS)

Antonio, Daniel; Mast, Daniel; Lavina, Barbara; Gofryk, Krzysztof

Uranium dioxide is the most commonly used nuclear fuel material in commercial reactors, while uranium nitride also has many thermal and physical properties that make it attractive for potential use in reactors. Both have a cubic fcc lattice structure at ambient conditions and transition to antiferromagnetic order at low temperature. UO2 is a Mott insulator that orders in a complex non-collinear 3k magnetic structure at about 30 K, while UN has appreciable conductivity and orders in a simpler 1k magnetic structure below 52 K. Both compounds are characterized by strong magneto-structural interactions, understanding of which is vital for modeling their thermo-physical properties. While UO2 and UN have been extensively studied at and above room temperature, little work has been done to directly study the structure of these materials at low temperatures where magnetic interactions are dominant. In the course of our systematic studies on magneto vibrational behavior of UO2 and UN, here we present our recent results of high pressure X-Ray Diffraction (up to 35 GPa) measured below the Neel temperature using synchrotron radiation. Work supported by the Department of Energy, Office of Basic Energy Sciences, Materials Sciences, and Engineering Division.

Physics of reactor safety. Quarterly report, January--March 1977. [LMFBR

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

None

1977-06-01

This report summarizes work done on reactor safety, Monte Carlo analysis of safety-related critical assembly experiments, and planning of DEMI safety-related critical experiments. Work on reactor core thermal-hydraulics is also included.

Fluidized bed gasification of industrial solid recovered fuels.

PubMed

Arena, Umberto; Di Gregorio, Fabrizio

2016-04-01

The study evaluates the technical feasibility of the fluidized bed gasification of three solid recovered fuels (SRFs), obtained as co-products of a recycling process. The SRFs were pelletized and fed to a pilot scale bubbling fluidized bed reactor, operated in gasification and co-gasification mode. The tests were carried out under conditions of thermal and chemical steady state, with a bed of olivine particles and at different values of equivalence ratio. The results provide a complete syngas characterization, in terms of its heating value and composition (including tars, particulates, and acid/basic pollutants) and of the chemical and physical characterization of bed material and entrained fines collected at the cyclone outlet. The feasibility of the fluidized bed gasification process of the different SRFs was evaluated with the support of a material and substance flow analysis, and a feedstock energy analysis. The results confirm the flexibility of fluidized bed reactor, which makes it one of the preferable technologies for the gasification of different kind of wastes, even in co-gasification mode. The fluidized bed gasification process of the tested SRFs appears technically feasible, yielding a syngas of valuable quality for energy applications in an appropriate plant configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

Low Energy Neutrino Physics at the Kuo-Sheng Reactor Laboratory in Taiwan

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

Lin, S.-T.

2006-11-17

A laboratory has been constructed by the TEXONO Collaboration at the Kuo-Sheng Reactor Power Plant in Taiwan to study low energy neutrino physics. A limit on the neutrino magnetic moment of {mu}{nu}({nu}-bare) < 7.2 x 10-11 {mu}B at 90% confidence level has been achieved from measurements with a high-purity germanium detector, as well as the electron neutrinos ({nu}{sub e}) produced from nuclear power reactors has been studied. Other research program at Kuo-Sheng are surveyed.

Impact of thorium based molten salt reactor on the closure of the nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Jaradat, Safwan Qasim Mohammad

Molten salt reactor (MSR) is one of six reactors selected by the Generation IV International Forum (GIF). The liquid fluoride thorium reactor (LFTR) is a MSR concept based on thorium fuel cycle. LFTR uses liquid fluoride salts as a nuclear fuel. It uses 232Th and 233U as the fertile and fissile materials, respectively. Fluoride salt of these nuclides is dissolved in a mixed carrier salt of lithium and beryllium (FLiBe). The objective of this research was to complete feasibility studies of a small commercial thermal LFTR. The focus was on neutronic calculations in order to prescribe core design parameter such as core size, fuel block pitch (p), fuel channel radius, fuel path, reflector thickness, fuel salt composition, and power. In order to achieve this objective, the applicability of Monte Carlo N-Particle Transport Code (MCNP) to MSR modeling was verified. Then, a prescription for conceptual small thermal reactor LFTR and relevant calculations were performed using MCNP to determine the main neutronic parameters of the core reactor. The MCNP code was used to study the reactor physics characteristics for the FUJI-U3 reactor. The results were then compared with the results obtained from the original FUJI-U3 using the reactor physics code SRAC95 and the burnup analysis code ORIPHY2. The results were comparable with each other. Based on the results, MCNP was found to be a reliable code to model a small thermal LFTR and study all the related reactor physics characteristics. The results of this study were promising and successful in demonstrating a prefatory small commercial LFTR design. The outcome of using a small core reactor with a diameter/height of 280/260 cm that would operate for more than five years at a power level of 150 MWth was studied. The fuel system 7LiF - BeF2 - ThF4 - UF4 with a (233U/ 232Th) = 2.01 % was the candidate fuel for this reactor core.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

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

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF 2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. Themore » objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.« less

Micro-reactors for characterization of nanostructure-based sensors.

PubMed

Savu, R; Silveira, J V; Flacker, A; Vaz, A R; Joanni, E; Pinto, A C; Gobbi, A L; Santos, T E A; Rotondaro, A L P; Moshkalev, S A

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO(2) substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10(-5) and 10(-1) mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Micro-reactors for characterization of nanostructure-based sensors

NASA Astrophysics Data System (ADS)

Savu, R.; Silveira, J. V.; Flacker, A.; Vaz, A. R.; Joanni, E.; Pinto, A. C.; Gobbi, A. L.; Santos, T. E. A.; Rotondaro, A. L. P.; Moshkalev, S. A.

2012-05-01

Fabrication and testing of micro-reactors for the characterization of nanosensors is presented in this work. The reactors have a small volume (100 μl) and are equipped with gas input/output channels. They were machined from a single piece of kovar in order to avoid leaks in the system due to additional welding. The contact pins were electrically insulated from the body of the reactor using a borosilicate sealing glass and the reactor was hermetically sealed using a lid and an elastomeric o-ring. One of the advantages of the reactor lies in its simple assembly and ease of use with any vacuum/gas system, allowing the connection of more than one device. Moreover, the lid can be modified in order to fit a window for in situ optical characterization. In order to prove its versatility, carbon nanotube-based sensors were tested using this micro-reactor. The devices were fabricated by depositing carbon nanotubes over 1 μm thick gold electrodes patterned onto Si/SiO2 substrates. The sensors were tested using oxygen and nitrogen atmospheres, in the pressure range between 10-5 and 10-1 mbar. The small chamber volume allowed the measurement of fast sensor characteristic times, with the sensors showing good sensitivity towards gas and pressure as well as high reproducibility.

Advanced 3D Characterization and Reconstruction of Reactor Materials FY16 Final Report

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

Fromm, Bradley; Hauch, Benjamin; Sridharan, Kumar

2016-12-01

A coordinated effort to link advanced materials characterization methods and computational modeling approaches is critical to future success for understanding and predicting the behavior of reactor materials that operate at extreme conditions. The difficulty and expense of working with nuclear materials have inhibited the use of modern characterization techniques on this class of materials. Likewise, mesoscale simulation efforts have been impeded due to insufficient experimental data necessary for initialization and validation of the computer models. The objective of this research is to develop methods to integrate advanced materials characterization techniques developed for reactor materials with state-of-the-art mesoscale modeling and simulationmore » tools. Research to develop broad-ion beam sample preparation, high-resolution electron backscatter diffraction, and digital microstructure reconstruction techniques; and methods for integration of these techniques into mesoscale modeling tools are detailed. Results for both irradiated and un-irradiated reactor materials are presented for FY14 - FY16 and final remarks are provided.« less

Thorium and Molten Salt Reactors: Essential Questions for Classroom Discussions

NASA Astrophysics Data System (ADS)

DiLisi, Gregory A.; Hirsch, Allison; Murray, Meredith; Rarick, Richard

2018-04-01

A little-known type of nuclear reactor called the "molten salt reactor" (MSR), in which nuclear fuel is dissolved in a liquid carrier salt, was proposed in the 1940s and developed at the Oak Ridge National Laboratory in the 1960s. Recently, the MSR has generated renewed interest as a remedy for the drawbacks associated with conventional uranium-fueled light-water reactors (LWRs) in use today. Particular attention has been given to the "thorium molten salt reactor" (TMSR), an MSR engineered specifically to use thorium as its fuel. The purpose of this article is to encourage the TPT community to incorporate discussions of MSRs and the thorium fuel cycle into courses such as "Physics and Society" or "Frontiers of Physics." With this in mind, we piloted a pedagogical approach with 27 teachers in which we described the underlying physics of the TMSR and posed five essential questions for classroom discussions. We assumed teachers had some preexisting knowledge of nuclear reactions, but such prior knowledge was not necessary for inclusion in the classroom discussions. Overall, our material was perceived as a real-world example of physics, fit into a standards-based curriculum, and filled a need in the teaching community for providing unbiased references of alternative energy technologies.

Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

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

Merzari, E.; Shemon, E. R.; Yu, Y. Q.

This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models ofmore » a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.« less

Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

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

Heeger, Karsten M.

2014-09-13

This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zeromore » $$\\theta_{13}$$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.« less

From the first nuclear power plant to fourth-generation nuclear power installations [on the 60th anniversary of the World's First nuclear power plant

NASA Astrophysics Data System (ADS)

Rachkov, V. I.; Kalyakin, S. G.; Kukharchuk, O. F.; Orlov, Yu. I.; Sorokin, A. P.

2014-05-01

Successful commissioning in the 1954 of the World's First nuclear power plant constructed at the Institute for Physics and Power Engineering (IPPE) in Obninsk signaled a turn from military programs to peaceful utilization of atomic energy. Up to the decommissioning of this plant, the AM reactor served as one of the main reactor bases on which neutron-physical investigations and investigations in solid state physics were carried out, fuel rods and electricity generating channels were tested, and isotope products were bred. The plant served as a center for training Soviet and foreign specialists on nuclear power plants, the personnel of the Lenin nuclear-powered icebreaker, and others. The IPPE development history is linked with the names of I.V. Kurchatov, A.I. Leipunskii, D.I. Blokhintsev, A.P. Aleksandrov, and E.P. Slavskii. More than 120 projects of various nuclear power installations were developed under the scientific leadership of the IPPE for submarine, terrestrial, and space applications, including two water-cooled power units at the Beloyarsk NPP in Ural, the Bilibino nuclear cogeneration station in Chukotka, crawler-mounted transportable TES-3 power station, the BN-350 reactor in Kazakhstan, and the BN-600 power unit at the Beloyarsk NPP. Owing to efforts taken on implementing the program for developing fast-neutron reactors, Russia occupied leading positions around the world in this field. All this time, IPPE specialists worked on elaborating the principles of energy supertechnologies of the 21st century. New large experimental installations have been put in operation, including the nuclear-laser setup B, the EGP-15 accelerator, the large physical setup BFS, the high-pressure setup SVD-2; scientific, engineering, and technological schools have been established in the field of high- and intermediate-energy nuclear physics, electrostatic accelerators of multicharge ions, plasma processes in thermionic converters and nuclear-pumped lasers, physics of compact nuclear reactors and radiation protection, thermal physics, physical chemistry and technology of liquid metal coolants, and physics of radiation-induced defects, and radiation materials science. The activity of the institute is aimed at solving matters concerned with technological development of large-scale nuclear power engineering on the basis of a closed nuclear fuel cycle with the use of fast-neutron reactors (referred to henceforth as fast reactors), development of innovative nuclear and conventional technologies, and extension of their application fields.

Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

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

Flach, G.P.

2000-02-11

A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data upmore » through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.« less

Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

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

Bond, Leonard J.; Bowler, John R.

The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-servicemore » inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO 3-xPbTiO 3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.« less

Coupled reactors analysis: New needs and advances using Monte Carlo methodology

DOE PAGES

Aufiero, M.; Palmiotti, G.; Salvatores, M.; ...

2016-08-20

Coupled reactors and the coupling features of large or heterogeneous core reactors can be investigated with the Avery theory that allows a physics understanding of the main features of these systems. However, the complex geometries that are often encountered in association with coupled reactors, require a detailed geometry description that can be easily provided by modern Monte Carlo (MC) codes. This implies a MC calculation of the coupling parameters defined by Avery and of the sensitivity coefficients that allow further detailed physics analysis. The results presented in this paper show that the MC code SERPENT has been successfully modifed tomore » meet the required capabilities.« less

ATOMIC PHYSICS, AN AUTOINSTRUCTIONAL PROGRAM, VOLUME 4, SUPPLEMENT.

ERIC Educational Resources Information Center

DETERLINE, WILLIAM A.; KLAUS, DAVID J.

THE AUTOINSTRUCTIONAL MATERIALS IN THIS TEXT WERE PREPARED FOR USE IN AN EXPERIMENTAL STUDY, OFFERING SELF-TUTORING MATERIAL FOR LEARNING ATOMIC PHYSICS. THE TOPICS COVERED ARE (1) RADIATION USES AND NUCLEAR FISSION, (2) NUCLEAR REACTORS, (3) ENERGY FROM NUCLEAR REACTORS, (4) NUCLEAR EXPLOSIONS AND FUSION, (5) A COMPREHENSIVE REVIEW, AND (6) A…

76 FR 48184 - Exelon Nuclear, Peach Bottom Atomic Power Station, Unit 1; Exemption From Certain Security...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-08

... nuclear reactor facility. PBAPS Unit 1 was a high-temperature, gas-cooled reactor that was operated from... the safeguards contingency plan.'' Part 73 of 10 CFR, ``Physical Protection of Plant and Materials... physical protection system which will have capabilities for the protection of special nuclear material at...

Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

ScienceCinema

None

2018-01-16

From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

ERIC Educational Resources Information Center

Malkawi, Salaheddin; Al-Araidah, Omar

2013-01-01

Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

The Virtual Environment for Reactor Applications (VERA): Design and architecture

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

Turner, John A.; Clarno, Kevin; Sieger, Matt

VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL), the first DOE Hub, which was established in July 2010 for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both amore » software and a numerical perspective, along with the goals and constraints that drove the major design decisions and their implications. As a result, we explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the application of VERA tools for a variety of challenging problems within the nuclear industry.« less

The Virtual Environment for Reactor Applications (VERA): Design and architecture

DOE PAGES

Turner, John A.; Clarno, Kevin; Sieger, Matt; ...

2016-09-08

VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL), the first DOE Hub, which was established in July 2010 for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both amore » software and a numerical perspective, along with the goals and constraints that drove the major design decisions and their implications. As a result, we explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the application of VERA tools for a variety of challenging problems within the nuclear industry.« 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).

Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

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

Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou

2016-06-01

As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterizationmore » of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.« less

Characterization of alternative electric generation technologies for the SPS comparative assessment. Volume 1: Summary of central station technologies

NASA Astrophysics Data System (ADS)

1980-08-01

The technologies selected for the detailed characterization were: solar technology; terrestrial photovoltaic (200 MWe); coal technologies; conventional high sulfur coal combustion with advanced fine gas desulfurization (1250 MWe), and open cycle gas turbine combined cycle plant with low Btu gasifier (1250 MWe); and nuclear technologies: conventional light water reactor (1250 MWe), liquid metal fast breeder reactor (1250 MWe), and magnetic fusion reactor (1320 MWe). A brief technical summary of each power plant design is given.

Characterization of the Caliban and Prospero Critical Assemblies Neutron Spectra for Integral Measurements Experiments

NASA Astrophysics Data System (ADS)

Casoli, P.; Authier, N.; Jacquet, X.; Cartier, J.

2014-04-01

Caliban and Prospero are two highly enriched uranium metallic core reactors operated on the CEA Center of Valduc. These critical assemblies are suitable for integral experiments, such as fission yields measurements or perturbation measurements, which have been carried out recently on the Caliban reactor. Different unfolding methods, based on activation foils and fission chambers measurements, are used to characterize the reactor spectra and especially the Caliban spectrum, which is very close to a pure fission spectrum.

CONVERTING FROM BATCH TO CONTINUOUS INTENSIFIED PROCESSING IN THE STT? REACTOR

EPA Science Inventory

The fluid dynamics, the physical dimensions and characteristics of the reaction zones of continuous process intensification reactors are often quite different from those of the batch reactors they replace. Understanding these differences is critical to the successful transit...

Characterization of Dust-Plasma Interactions In Non-Thermal Plasmas Under Low Pressure and the Atmospheric Pressure

NASA Astrophysics Data System (ADS)

Bilik, Narula

This dissertation research focuses on the experimental characterization of dust-plasma interactions at both low and atmospheric pressure. Its goal is to fill the knowledge gaps in (1) the fundamental research of low pressure dusty plasma electrons, which mainly relied on models with few experimental results; and (2) the nanoparticle synthesis process in atmospheric pressure uniform glow plasmas (APGDs), which is largely unexplored in spite of the economical advantage of APGDs in nanotechnology. The low pressure part of the dissertation research involves the development of a complete diagnostic process for an argon-siline capacitively-coupled RF plasma. The central part of the diagnostic process is the Langmuir probe measurement of the electron energy probability function (EEPF) in a dusty plasma, which has never been measured before. This is because the dust particles in the plasma cause severe probe surface contamination and consequently distort the measurement. This problem is solved by adding a solenoid-actuated shield structure to the Langmuir probe, which physically protects the Langmuir probe from the dust particle deposition to ensure reliable EEPF measurements. The dusty plasma EEPFs are characterized by lower electron density and higher electron temperature accompanied by a drop in the low energy electron population. The Langmuir probe measurement is complemented with other characterizations including the capacitive probe measurement, power measurement, and dust particle collection. The complete diagnostic process then gives a set of local plasma parameters as well as the details of the dust-electron interactions reflected in the EEPFs. This set of data serves as input for an analytical model of nanoparticle charging to yield the time evolution of nanoparticle size and charge in the dusty plasma. The atmospheric pressure part of the dissertation focuses on the design and development of an APGD for zinc oxide nanocrystal synthesis. One of the main difficulties in maintaining an APGD is ensuring its uniformity over large discharge volume. By examining past atmospheric pressure plasma reactor designs and looking into the details of the atmospheric pressure gas breakdown mechanism, three design features are proposed to ensure the APGD uniformity. These include the use of a dielectric barrier and the RF driving frequency, as well as a pre-ionization technique achieved by having a non-uniform gap spacing in a capacitively-coupled concentric cylinder reactor. The resulting APGD reactor operates stably in the abnormal glow regime using either helium or argon as the carrier gas. Diethylzinc (DEZ) and oxygen precursors are injected into the APGD to form zinc oxide nanocrystals. The physical and optical properties of these nanocrystals are characterized, and the system parameters that impact the nanoparticle size and deposition rate are identified.

Comprehensive Experiments on Subcritical Assemblies of Cascade Reactor Systems

NASA Astrophysics Data System (ADS)

Zavyalov, N. V.; Il'kaev, R. I.; Kolesov, V. F.; Ivanin, I. A.; Zhitnik, A. K.; Kuvshinov, M. I.; Nefedov, Yu. Ya.; Punin, V. T.; Tel'nov, A. V.; Khoruzhi, V. Kh.

2017-12-01

Cascade reactors attract particular attention because of their capability of improving the parameters of pulsed reactors and achieving the feasibility of electronuclear facilities. The paper presents the results of three series of experiments on uranium-neptunium cascade assemblies at the Institute of Nuclear and Radiation Physics of the All-Russian Research Institute of Experimental Physics conducted in 2003-2004. The experiments confirmed theoretical conclusions on positive properties of cascade blankets and effectiveness of using neptunium-237 as a means of creating a one-sided connection between the sections.

Modeling of a Flooding Induced Station Blackout for a Pressurized Water Reactor Using the RISMC Toolkit

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

Mandelli, Diego; Prescott, Steven R; Smith, Curtis L

2011-07-01

In the Risk Informed Safety Margin Characterization (RISMC) approach we want to understand not just the frequency of an event like core damage, but how close we are (or are not) to key safety-related events and how might we increase our safety margins. The RISMC Pathway uses the probabilistic margin approach to quantify impacts to reliability and safety by coupling both probabilistic (via stochastic simulation) and mechanistic (via physics models) approaches. This coupling takes place through the interchange of physical parameters and operational or accident scenarios. In this paper we apply the RISMC approach to evaluate the impact of amore » power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., system activation) and to perform statistical analyses (e.g., run multiple RELAP-7 simulations where sequencing/timing of events have been changed according to a set of stochastic distributions). By using the RISMC toolkit, we can evaluate how power uprate affects the system recovery measures needed to avoid core damage after the PWR lost all available AC power by a tsunami induced flooding. The simulation of the actual flooding is performed by using a smooth particle hydrodynamics code: NEUTRINO.« less

Nuclear Reactor Safety--The APS Submits its Report

ERIC Educational Resources Information Center

Physics Today, 1975

1975-01-01

Presents the summary section of the American Physical Society (APS) report on the safety features of the light-water reactor, reviews the design, construction, and operation of a reactor and outlines the primary engineered safety features. Summarizes the major recommendations of the study group. (GS)

Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

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

Sterbentz, James William; Werner, James Elmer; McKellar, Michael George

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heatmore » exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.« less

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)

A Special Topic From Nuclear Reactor Dynamics for the Undergraduate Physics Curriculum

ERIC Educational Resources Information Center

Sevenich, R. A.

1977-01-01

Presents an intuitive derivation of the point reactor equations followed by formulation of equations for inverse and direct kinetics which are readily programmed on a digital computer. Suggests several computer simulations involving the effect of control rod motion on reactor power. (MLH)

An Introduction to the Issues

ERIC Educational Resources Information Center

Primack, Joel

1975-01-01

The reactor safety controversy is reviewed in light of the United States Atomic Energy Commission's Reactor Safety Study and the Report to the American Physical Society by the Study Group on Light Water Reactor Safety. Areas of agreement and disagreement are identified and implications for national policy are explored. (BT)

A useful observable for estimating keff in fast subcritical systems

NASA Astrophysics Data System (ADS)

Saracco, Paolo; Borreani, Walter; Chersola, Davide; Lomonaco, Guglielmo; Ricco, Gianni; Ripani, Marco

2017-09-01

The neutron multiplication factor keff is a key quantity to characterize subcritical neutron multiplying devices and for understanting their physical behaviour, being related to the fundamental eigenvalue of Boltzmann transport equation. Both the maximum available power - and all quantities related to it, like, e.g. the effectiveness in burning nuclear wastes - as well as reactor kinetics and dynamics depend on keff. Nevertheless, keff is not directly measurable and its determination results from the solution of an inverse problem: minimizing model dependence of the solution for keff then becomes a critical issue, relevant both for practical and theoretical reasons.

Self isolating high frequency saturable reactor

DOEpatents

Moore, James A.

1998-06-23

The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

Photocatalytic production of hydrogen from fixed titanium dioxide thin film

NASA Astrophysics Data System (ADS)

Okoye, Njideka Helen

This thesis is focused on further developing of an efficient method for the photocatalytic hydrogen production. The research aimed to use thin films deposited with TiO2 and doped with Pt in order to substitute slurry solutions that are currently being used. A new depositing experimental approach to manufacture the thin films was proposed and tested for both physical properties and chemical reactivity. Therefore, the experiment was designed into two parts: The first part was on the manufacturing and the physical characterization of titanium dioxide deposited on glass surfaces and the second part was focused on the ability of the thin film to produce hydrogen. For the second part, a photochemical reactor vessel was used to properly place the glass slides to UV-irradiation. This was yielded by a mercury lamp located at the centre of the reactor. The thesis is organized into five different chapters including introduction, literature review, characterization of TiO2 coated surface, experimental design and hydrogen production, finally conclusive observations and future work. Hydrogen production by photodecomposition of water into H2 and O2 has a very low efficiency due to rapid reverse reaction and, as mentioned above, it usually requires a slurry type of solution. This needs additional processing steps such as filtration and recycling of particles. Therefore, it is important to develop an efficient process for hydrogen production. TiO2 coated surfaces could be an excellent technological alternative. In this study, a sol-gel method was used to produce a transparent TiO 2 thin film which was deposited on a glass substrate by using a new coating technique introduced in this work for H2 production. The TiO2 deposited film on a glass substrate by using the spraying method of coating was characterized for physical analysis (surface characteristics, size of nanoparticles and distribution, etc.) by using X-Ray Diffraction (XRD), Scanning Electron Microscope (SEM), Transmission Electron Microscope (TEM) and UV-Visible optical fiber spectrophotometer. Platinum was deposited on the coated thin film by adsorption from aqueous solutions containing Na 2PtCl4 followed by calcination at 500o C. The chemical reactivity of the new coated thin film for H2 production was tested by examining the effect of different ethanol concentrations and surface area available to hydrogen production rate by using a mercury lamp in a bench scale photo reactor with ethanol and water as the reactants. It was observed over a period of two hr interval that increase in ethanol concentration investigated in this work showed substantial increase in the hydrogen production rate as well as when increasing the surface area.

Dependency of the Reynolds number on the water flow through the perforated tube

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

Závodný, Zdenko, E-mail: zdenko.zavodny@stuba.sk; Bereznai, Jozef, E-mail: jozef.bereznai@stuba.sk; Urban, František

Safe and effective loading of nuclear reactor fuel assemblies demands qualitative and quantitative analysis of the relationship between the coolant temperature in the fuel assembly outlet, measured by the thermocouple, and the mean coolant temperature profile in the thermocouple plane position. It is not possible to perform the analysis directly in the reactor, so it is carried out using measurements on the physical model, and the CFD fuel assembly coolant flow models. The CFD models have to be verified and validated in line with the temperature and velocity profile obtained from the measurements of the cooling water flowing in themore » physical model of the fuel assembly. Simplified physical model with perforated central tube and its validated CFD model serve to design of the second physical model of the fuel assembly of the nuclear reactor VVER 440. Physical model will be manufactured and installed in the laboratory of the Institute of Energy Machines, Faculty of Mechanical Engineering of the Slovak University of Technology in Bratislava.« less

ReactorHealth Physics operations at the NIST center for neutron research.

PubMed

Johnston, Thomas P

2015-02-01

Performing health physics and radiation safety functions under a special nuclear material license and a research and test reactor license at a major government research and development laboratory encompasses many elements not encountered by industrial, general, or broad scope licenses. This article reviews elements of the health physics and radiation safety program at the NIST Center for Neutron Research, including the early history and discovery of the neutron, applications of neutron research, reactor overview, safety and security of radiation sources and radioactive material, and general health physics procedures. These comprise precautions and control of tritium, training program, neutron beam sample processing, laboratory audits, inventory and leak tests, meter calibration, repair and evaluation, radioactive waste management, and emergency response. In addition, the radiation monitoring systems will be reviewed including confinement building monitoring, ventilation filter radiation monitors, secondary coolant monitors, gaseous fission product monitors, gas monitors, ventilation tritium monitor, and the plant effluent monitor systems.

High-performance recombinant protein production with Escherichia coli in continuously operated cascades of stirred-tank reactors.

PubMed

Schmideder, Andreas; Weuster-Botz, Dirk

2017-07-01

The microbial expression of intracellular, recombinant proteins in continuous bioprocesses suffers from low product concentrations. Hence, a process for the intracellular production of photoactivatable mCherry with Escherichia coli in a continuously operated cascade of two stirred-tank reactors was established to separate biomass formation (first reactor) and protein expression (second reactor) spatially. Cascades of miniaturized stirred-tank reactors were implemented, which enable the 24-fold parallel characterization of cascade processes and the direct scale-up of results to the liter scale. With PAmCherry concentrations of 1.15 g L -1 cascades of stirred-tank reactors improved the process performance significantly compared to production processes in chemostats. In addition, an optimized fed-batch process was outperformed regarding space-time yield (149 mg L -1  h -1 ). This study implicates continuous cascade processes to be a promising alternative to fed-batch processes for microbial protein production and demonstrates that miniaturized stirred-tank reactors can reduce the timeline and costs for cascade process characterization.

40 CFR 60.701 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... Synthetic Organic Chemical Manufacturing Industry (SOCMI) Reactor Processes § 60.701 Definitions. As used in... means any noncontinuous reactor process that is not characterized by steady-state conditions and in.... Reactor processes are unit operations in which one or more chemicals, or reactants other than air, are...

Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

NASA Astrophysics Data System (ADS)

Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

2017-01-01

In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

RH-TRU Waste Characterization by Acceptable Knowledge at the Idaho National Engineering and Environmental Laboratory

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

Schulz, C.; Givens, C.; Bhatt, R.

2003-02-24

Idaho National Engineering and Environmental Laboratory (INEEL) is conducting an effort to characterize approximately 620 drums of remote-handled (RH-) transuranic (TRU) waste currently in its inventory that were generated at the Argonne National Laboratory-East (ANL-E) Alpha Gamma Hot Cell Facility (AGHCF) between 1971 and 1995. The waste was generated at the AGHCF during the destructive examination of irradiated and unirradiated fuel pins, targets, and other materials from reactor programs at ANL-West (ANL-W) and other Department of Energy (DOE) reactors. In support of this effort, Shaw Environmental and Infrastructure (formerly IT Corporation) developed an acceptable knowledge (AK) collection and management programmore » based on existing contact-handled (CH)-TRU waste program requirements and proposed RH-TRU waste program requirements in effect in July 2001. Consistent with Attachments B-B6 of the Waste Isolation Pilot Plant (WIPP) Hazardous Waste Facility Permit (HWFP) and th e proposed Class 3 permit modification (Attachment R [RH-WAP] of this permit), the draft AK Summary Report prepared under the AK procedure describes the waste generating process and includes determinations in the following areas based on AK: physical form (currently identified at the Waste Matrix Code level); waste stream delineation; applicability of hazardous waste numbers for hazardous waste constituents; and prohibited items. In addition, the procedure requires and the draft summary report contains information supporting determinations in the areas of defense relationship and radiological characterization.« less

Small Reactor for Deep Space Exploration

ScienceCinema

none,

2018-06-06

This is the first demonstration of a space nuclear reactor system to produce electricity in the United States since 1965, and an experiment demonstrated the first use of a heat pipe to cool a small nuclear reactor and then harvest the heat to power a Stirling engine at the Nevada National Security Site's Device Assembly Facility confirms basic nuclear reactor physics and heat transfer for a simple, reliable space power system.

Model Studies on the Effectiveness of MBBR Reactors for the Restoration of Small Water Reservoirs

NASA Astrophysics Data System (ADS)

Nowak, Agata; Mazur, Robert; Panek, Ewa; Chmist, Joanna

2018-02-01

The authors present the Moving Bed Biofilm Reactor (MBBR) model with a quasi-continuous flow for small water reservoir restoration, characterized by high concentrations of organic pollutants. To determine the efficiency of wastewater treatment the laboratory analysis of physic-chemical parameters were conducted for the model on a semi-technical scale of 1:3. Wastewater treatment process was carried out in 24 h for 1 m3 for raw sewage. The startup period was 2 weeks for all biofilters (biological beds). Approximately 50% reduction in COD and BOD5 was obtained on average for the studied bioreactors. Significant improvements were achieved in theclarity of the treated wastewater, with the reduction of suspension by 60%. The oxygen profile has improved significantly in 7 to 9 hours of the process, and a diametric reduction in the oxidative reduction potential was recorded. A preliminary model of biological treatment effectiveness was determined based on the conducted studies. In final stages, the operation mode was set in real conditions of polluted water reservoirs.

Neutrino Physics at Drexel

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

Lane, Charles; Dolinski, Michelle; Neilson, Russell

Our primary goal is to improve the understanding of the properties and interactions of neutrinos. We are pursuing this by means of the DUNE long-baseline and PROSPECT short-baseline neutrino experiments. For DUNE, a neutrino beam from Fermilab will be detected at the SURF facility in South Dakota, with the aim of determining the neutrino mass hierarchy (the mass ordering of neutrino flavors), and a measurement or limit on CP-violation via neutrinos. Our near-term experimental goal is to improve the characterization of the neutrino beam by measurements of muons produced as a byproduct of neutrino beam generation, to quantify the beammore » composition and flux. The short-range neutrino program has the aim of using the HFIR reactor at Oak Ridge as a neutrino source, with a detector placed nearby to find if there are short-distance oscillations to sterile neutrino flavors, and to resolve the 'reactor neutrino spectral anomaly' which has shown up as an unexplained 'bump' in the neutrino energy spectrum in recent experiments.« less

Characterization of Used Nuclear Fuel with Multivariate Analysis for Process Monitoring

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

Dayman, Kenneth J.; Coble, Jamie B.; Orton, Christopher R.

2014-01-01

The Multi-Isotope Process (MIP) Monitor combines gamma spectroscopy and multivariate analysis to detect anomalies in various process streams in a nuclear fuel reprocessing system. Measured spectra are compared to models of nominal behavior at each measurement location to detect unexpected changes in system behavior. In order to improve the accuracy and specificity of process monitoring, fuel characterization may be used to more accurately train subsequent models in a full analysis scheme. This paper presents initial development of a reactor-type classifier that is used to select a reactor-specific partial least squares model to predict fuel burnup. Nuclide activities for prototypic usedmore » fuel samples were generated in ORIGEN-ARP and used to investigate techniques to characterize used nuclear fuel in terms of reactor type (pressurized or boiling water reactor) and burnup. A variety of reactor type classification algorithms, including k-nearest neighbors, linear and quadratic discriminant analyses, and support vector machines, were evaluated to differentiate used fuel from pressurized and boiling water reactors. Then, reactor type-specific partial least squares models were developed to predict the burnup of the fuel. Using these reactor type-specific models instead of a model trained for all light water reactors improved the accuracy of burnup predictions. The developed classification and prediction models were combined and applied to a large dataset that included eight fuel assembly designs, two of which were not used in training the models, and spanned the range of the initial 235U enrichment, cooling time, and burnup values expected of future commercial used fuel for reprocessing. Error rates were consistent across the range of considered enrichment, cooling time, and burnup values. Average absolute relative errors in burnup predictions for validation data both within and outside the training space were 0.0574% and 0.0597%, respectively. The errors seen in this work are artificially low, because the models were trained, optimized, and tested on simulated, noise-free data. However, these results indicate that the developed models may generalize well to new data and that the proposed approach constitutes a viable first step in developing a fuel characterization algorithm based on gamma spectra.« less

Reactor monitoring using antineutrino detectors

NASA Astrophysics Data System (ADS)

Bowden, N. S.

2011-08-01

Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactor as part of International Atomic Energy Agency (IAEA) and/or other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway worldwide.

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.

Plasma characterization studies for materials processing

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

Pfender, E.; Heberlein, J.

New applications for plasma processing of materials require a more detailed understanding of the fundamental processes occurring in the processing reactors. We have developed reactors offering specific advantages for materials processing, and we are using modeling and diagnostic techniques for the characterization of these reactors. The emphasis is in part set by the interest shown by industry pursuing specific plasma processing applications. In this paper we report on the modeling of radio frequency plasma reactors for use in materials synthesis, and on the characterization of the high rate diamond deposition process using liquid precursors. In the radio frequency plasma torchmore » model, the influence of specific design changes such as the location of the excitation coil on the enthalpy flow distribution is investigated for oxygen and air as plasma gases. The diamond deposition with liquid precursors has identified the efficient mass transport in form of liquid droplets into the boundary layer as responsible for high growth, and the chemical properties of the liquid for the film morphology.« less

Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

NASA Astrophysics Data System (ADS)

Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

2018-05-01

The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

Characterization of Metalorganic Chemical Vapor Deposition

NASA Technical Reports Server (NTRS)

Jesser, W. A.

1998-01-01

A series of experimental and numerical investigations to develop a more complete understanding of the reactive fluid dynamics of chemical vapor deposition were conducted. In the experimental phases of the effort, a horizontal CVD reactor configuration was used for the growth of InP at UVA and for laser velocimetry measurements of the flow fields in the reactor at LaRC. This horizontal reactor configuration was developed for the growth of III-V semiconductors and has been used by our research group in the past to study the deposition of both GaAs and InP. While the ultimate resolution of many of the heat and mass transport issues will require access to a reduced-gravity environment, the series of groundbased research makes direct contributions to this area while attempting to answer the design questions for future experiments of how low must gravity be reduced and for how long must this gravity level be maintained to make the necessary measurements. It is hoped that the terrestrial experiments will be useful for the design of future microgravity experiments which likely will be designed to employ a core set of measurements for applications in the microgravity environment such as HOLOC, the Fluid Physics/Dynamics Facility, or the Schlieren photography, the Laser Imaging Velocimetry and the Laser Doppler Velocimetry instruments under development for the Advanced Fluids Experiment Module.

Effects of water vapor pretreatment time and reaction temperature on CO(2) capture characteristics of a sodium-based solid sorbent in a bubbling fluidized-bed reactor.

PubMed

Seo, Yongwon; Jo, Sung-Ho; Ryu, Chong Kul; Yi, Chang-Keun

2007-10-01

CO(2) capture from flue gas using a sodium-based solid sorbent was investigated in a bubbling fluidized-bed reactor. Carbonation and regeneration temperature on CO(2) removal was determined. The extent of the chemical reactivity after carbonation or regeneration was characterized via (13)C NMR. In addition, the physical properties of the sorbent such as pore size, pore volume, and surface area after carbonation or regeneration were measured by gas adsorption method (BET). With water vapor pretreatment, near complete CO(2) removal was initially achieved and maintained for about 1-2min at 50 degrees C with 2s gas residence time, while without proper water vapor pretreatment CO(2) removal abruptly decreased from the beginning. Carbonation was effective at the lower temperature over the 50-70 degrees C temperature range, while regeneration more effective at the higher temperature over the 135-300 degrees C temperature range. To maintain the initial 90% CO(2) removal, it would be necessary to keep the regeneration temperature higher than about 135 degrees C. The results obtained in this study can be used as basic data for designing and operating a large scale CO(2) capture process with two fluidized-bed reactors.

A Passive System Reliability Analysis for a Station Blackout

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

Brunett, Acacia; Bucknor, Matthew; Grabaskas, David

2015-05-03

The latest iterations of advanced reactor designs have included increased reliance on passive safety systems to maintain plant integrity during unplanned sequences. While these systems are advantageous in reducing the reliance on human intervention and availability of power, the phenomenological foundations on which these systems are built require a novel approach to a reliability assessment. Passive systems possess the unique ability to fail functionally without failing physically, a result of their explicit dependency on existing boundary conditions that drive their operating mode and capacity. Argonne National Laboratory is performing ongoing analyses that demonstrate various methodologies for the characterization of passivemore » system reliability within a probabilistic framework. Two reliability analysis techniques are utilized in this work. The first approach, the Reliability Method for Passive Systems, provides a mechanistic technique employing deterministic models and conventional static event trees. The second approach, a simulation-based technique, utilizes discrete dynamic event trees to treat time- dependent phenomena during scenario evolution. For this demonstration analysis, both reliability assessment techniques are used to analyze an extended station blackout in a pool-type sodium fast reactor (SFR) coupled with a reactor cavity cooling system (RCCS). This work demonstrates the entire process of a passive system reliability analysis, including identification of important parameters and failure metrics, treatment of uncertainties and analysis of results.« less

Physical-chemical treatment of wastes: a way to close turnover of elements in LSS

NASA Astrophysics Data System (ADS)

Kudenko, Yu A.; Gribovskaya, I. V.; Zolotukhin, I. G.

2000-05-01

"Man-plants-physical-chemical unit" system designed for space stations or terrestrial ecohabitats to close steady-state mineral, water and gas exchange is proposed. The physical-chemical unit is to mineralize all inedible plant wastes and physiological human wastes (feces, urine, gray water) by electromagnetically activated hydrogen peroxide in an oxidation reactor. The final product is a mineralized solution containing all elements balanced for plants' requirements. The solution has been successfully used in experiments to grow wheat, beans and radish. The solution was reusable: the evaporated moisture was replenished by the phytotron condensate. Sodium salination of plants was precluded by evaporating reactor-mineralized urine to sodium saturation concentration to crystallize out NaCl which can be used as food for the crew. The remaining mineralized product was brought back for nutrition of plants. The gas composition of the reactor comprises O 2, N 2, CO 2, NH 3, H 2. At the reactor's output hydrogen and oxygen were catalyzed into water, NH 3 was converted in a water trap into NH 4 and used for nutrition of plants. A special accessory at the reactor's output may produce hydrogen peroxide from intrasystem water and gas which makes possible to close gas loops between LSS components.

Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

NASA Technical Reports Server (NTRS)

Butler, C.; Albright, D.

2007-01-01

Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

Fundamental interactions involving neutrons and neutrinos: reactor-based studies led by Petersburg Nuclear Physics Institute (National Research Centre 'Kurchatov Institute') [PNPI (NRC KI)

NASA Astrophysics Data System (ADS)

Serebrov, A. P.

2015-11-01

Neutrons of very low energy ( ˜ 10-7 eV), commonly known as ultracold, are unique in that they can be stored in material and magnetic traps, thus enhancing methodical opportunities to conduct precision experiments and to probe the fundamentals of physics. One of the central problems of physics, of direct relevance to the formation of the Universe, is the violation of time invariance. Experiments searching for the nonzero neutron electric dipole moment serve as a time invariance test, and the use of ultracold neutrons provides very high measurement precision. Precision neutron lifetime measurements using ultracold neutrons are extremely important for checking ideas on the early formation of the Universe. This paper discusses problems that arise in studies using ultracold neutrons. Also discussed are the currently highly topical problem of sterile neutrinos and the search for reactor antineutrino oscillations at distances of 6-12 meters from the reactor core. The field reviewed is being investigated at multiple facilities globally. The present paper mainly concentrates on the results of PNPI-led studies at WWR-M PNPI (Gatchina), ILL (Grenoble), and SM-3 (Dimitrovgrad) reactors, and also covers the results obtained during preparation for research at the PIK reactor which is under construction.

Nuclear power industry: Tendencies in the world and Ukraine

NASA Astrophysics Data System (ADS)

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

2007-11-01

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

Simulation of Watts Bar Unit 1 Initial Startup Tests with Continuous Energy Monte Carlo Methods

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

Godfrey, Andrew T; Gehin, Jess C; Bekar, Kursat B

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors* is developing a collection of methods and software products known as VERA, the Virtual Environment for Reactor Applications. One component of the testing and validation plan for VERA is comparison of neutronics results to a set of continuous energy Monte Carlo solutions for a range of pressurized water reactor geometries using the SCALE component KENO-VI developed by Oak Ridge National Laboratory. Recent improvements in data, methods, and parallelism have enabled KENO, previously utilized predominately as a criticality safety code, to demonstrate excellent capability and performance for reactor physics applications. The highlymore » detailed and rigorous KENO solutions provide a reliable nu-meric reference for VERAneutronics and also demonstrate the most accurate predictions achievable by modeling and simulations tools for comparison to operating plant data. This paper demonstrates the performance of KENO-VI for the Watts Bar Unit 1 Cycle 1 zero power physics tests, including reactor criticality, control rod worths, and isothermal temperature coefficients.« less

On Study of Application of Micro-reactor in Chemistry and Chemical Field

NASA Astrophysics Data System (ADS)

Zhang, Yunshen

2018-02-01

Serving as a micro-scale chemical reaction system, micro-reactor is characterized by high heat transfer efficiency and mass transfer, strictly controlled reaction time and good safety performance; compared with the traditional mixing reactor, it can effectively shorten reaction time by virtue of these advantages and greatly enhance the chemical reaction conversion rate. However, problems still exist in the process where micro-reactor is used for production in chemistry and chemical field, and relevant researchers are required to optimize and perfect the performance of micro-reactor. This paper analyzes specific application of micro-reactor in chemistry and chemical field.

Double Chooz and a history of reactor θ 13 experiments

DOE PAGES

Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

2016-04-11

This is a contribution paper from the Double Chooz (DC) experiment to the special issue of Nuclear Physics B on the topics of neutrino oscillations, celebrating the recent Nobel prize to Profs. T. Kajita and A.B. McDonald. DC is a reactor neutrino experiment which measures the last neutrino mixing angle θ 13. In addition, the DC group presented an indication of disappearance of the reactor neutrinos at a baseline of similar to 1 km for the first time in 2011 and is improving the measurement of θ 13. DC is a pioneering experiment of this research field. In accordance withmore » the nature of this special issue, physics and history of the reactor-θ 13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.« less

Double Chooz and a history of reactor θ 13 experiments

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

Suekane, Fumihiko; Junqueira de Castro Bezerra, Thiago

This is a contribution paper from the Double Chooz (DC) experiment to the special issue of Nuclear Physics B on the topics of neutrino oscillations, celebrating the recent Nobel prize to Profs. T. Kajita and A.B. McDonald. DC is a reactor neutrino experiment which measures the last neutrino mixing angle θ 13. In addition, the DC group presented an indication of disappearance of the reactor neutrinos at a baseline of similar to 1 km for the first time in 2011 and is improving the measurement of θ 13. DC is a pioneering experiment of this research field. In accordance withmore » the nature of this special issue, physics and history of the reactor-θ 13 experiments, as well as the Double Chooz experiment and its neutrino oscillation analyses, are reviewed.« less

Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

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

S. Bragg-Sitton; J. Bess; J. Werner

2011-09-01

Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al.,more » 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).« less

Summary of ORSphere critical and reactor physics measurements

NASA Astrophysics Data System (ADS)

Marshall, Margaret A.; Bess, John D.

2017-09-01

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

MeV per Nucleon Ion Irradiation of Nuclear Materials with High Energy Synchrotron X-ray Characterization

DOE PAGES

Pellin, M. J.; Yacout, Abdellatif M.; Mo, Kun; ...

2016-01-14

The combination of MeV/Nucleon ion irradiation (e.g. 133 MeV Xe) and high energy synchrotron x-ray characterization (e.g. at the Argonne Advanced Photon Source, APS) provides a powerful characterization method to understand radiation effects and to rapidly screen materials for the nuclear reactor environment. Ions in this energy range penetrate ~10 μm into materials. Over this range, the physical interactions vary (electronic stopping, nuclear stopping and added interstitials). Spatially specific x-ray (and TEM and nanoindentation) analysis allow individual quantification of these various effects. Hard x-rays provide the penetration depth needed to analyze even nuclear fuels. Here, this combination of synchrotron x-raymore » and MeV/Nucleon ion irradiation is demonstrated on U-Mo fuels. A preliminary look at HT-9 steels is also presented. We suggest that a hard x-ray facility with in situ MeV/nucleon irradiation capability would substantially accelerate the rate of discovery for extreme materials.« less

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.

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

Hanford Atomic Products Operation monthly report for February 1956

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

Not Available

1956-02-21

This is the monthly report for the Hanford Laboratories Operation, February, 1956. Metallurgy, reactors fuels, chemistry, dosimetry, separation processes, reactor technology financial activities, visits, biology operation, physics and instrumentation research, employee relations are discussed.

Hanford Laboratories monthly activities report, March 1964

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

Not Available

1964-04-15

The monthly report for the Hanford Laboratories Operation, March 1964. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research, and applied mathematics operation, and programming operations are discussed.

Hanford Laboratories Operation monthly activities report, September 1960

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

Not Available

1960-10-15

This is the monthly report for the Hanford Laboratories Operation, October, 1960. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

Hanford Laboratories monthly activities report, August 1963

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

Not Available

1963-09-16

This is the monthly report for the Hanford Laboratories Operation, August 1963. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

Hanford Laboratories Operation monthly activities report, November 1962

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

Not Available

1962-12-14

This is the monthly report for the Hanford Laboratories Operation, November 1962. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

A document review to characterize Atomic International SNAP fuels shipped to INEL 1966--1973

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

Wahnschaffe, S.D.; Lords, R.E.; Kneff, D.W.

1995-09-01

This report provides the results of a document search and review study to obtain information on the spent fuels for the following six Nuclear Auxiliary Power (SNAP) reactor cores now stored at the Idaho National Engineering Laboratory (INEL): SNAP-2 Experimental Reactor, SNAP-2 Development Reactor, SNAP-10A Ground Test Reactor, SNAP-8 Experimental Reactor, SNAP-8 Development Reactor, and Shield Test Reactor. The report also covers documentation on SNAP fuel materials from four in-pile materials tests: NAA-82-1, NAA-115-2, NAA-117-1, and NAA-121. Pieces of these fuel materials are also stored at INEL as part of the SNAP fuel shipments.

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL

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

D. Kokkinos

2005-04-28

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophymore » on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.« less

Nuclear physics research operation. Monthly report, November 1958

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

Faulkner, J.E.

1958-12-10

This report is a summary of projects worked on in support of the production reactors at Hanford. The projects include criticality studies, from tasks associated with fuel element reprocessing to shipments of slightly enriched uranium. They include studies of neutron cross sections for different reactions and neutron flux measurements in different reactor locations, as well as design studies for future reactor projects.

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

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.

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

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

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Hanford Laboratories monthly activities report, February 1964

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

Not Available

1964-03-16

This is the monthly report for the Hanford Laboratories Operation, February, 1964. Reactor fuels, chemistry, dosimetry, separation process, reactor technology financial activities, biology operation, physics and instrumentation research, employee relations, applied mathematics, programming, and radiation protection are discussed.

Hanford Atomic Products Operation monthly report for June 1955

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

Not Available

1955-07-28

This is the monthly report for the Hanford Atomic Products Operation, June, 1955. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

Hanford Atomic Products Operation monthly report, January 1956

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

Not Available

1956-02-24

This is the monthly report for the Hanford Atomic Laboratories Products Operation, February, 1956. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, visits, biology operation, physics and instrumentation research, and employee relations are discussed.

Introduction to D-He(3) fusion reactors

NASA Technical Reports Server (NTRS)

Vlases, G. C.; Steinhauer, L. C.

1989-01-01

A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

RELAP-7 Software Verification and Validation Plan

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

Smith, Curtis L.; Choi, Yong-Joon; Zou, Ling

This INL plan comprehensively describes the software for RELAP-7 and documents the software, interface, and software design requirements for the application. The plan also describes the testing-based software verification and validation (SV&V) process—a set of specially designed software models used to test RELAP-7. The RELAP-7 (Reactor Excursion and Leak Analysis Program) code is a nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on the INL’s modern scientific software development framework – MOOSE (Multi-Physics Object-Oriented Simulation Environment). The overall design goal of RELAP-7 is to take advantage of the previous thirty yearsmore » of advancements in computer architecture, software design, numerical integration methods, and physical models. The end result will be a reactor systems analysis capability that retains and improves upon RELAP5’s capability and extends the analysis capability for all reactor system simulation scenarios.« less

Introduction to D-He(3) fusion reactors

NASA Astrophysics Data System (ADS)

Vlases, G. C.; Steinhauer, L. C.

1989-07-01

A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

Virtual environments simulation in research reactor

NASA Astrophysics Data System (ADS)

Muhamad, Shalina Bt. Sheik; Bahrin, Muhammad Hannan Bin

2017-01-01

Virtual reality based simulations are interactive and engaging. It has the useful potential in improving safety training. Virtual reality technology can be used to train workers who are unfamiliar with the physical layout of an area. In this study, a simulation program based on the virtual environment at research reactor was developed. The platform used for virtual simulation is 3DVia software for which it's rendering capabilities, physics for movement and collision and interactive navigation features have been taken advantage of. A real research reactor was virtually modelled and simulated with the model of avatars adopted to simulate walking. Collision detection algorithms were developed for various parts of the 3D building and avatars to restrain the avatars to certain regions of the virtual environment. A user can control the avatar to move around inside the virtual environment. Thus, this work can assist in the training of personnel, as in evaluating the radiological safety of the research reactor facility.

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.

Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

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

C. Fiorina; N. E. Stauff; F. Franceschini

2013-12-01

The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associatedmore » with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.« less

Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

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

Nigg, David W.; Nielsen, Joseph W.; Norman, Daren R.

The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be wellmore » outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.« less

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

Characterization and treatment of Denizli landfill leachate using anaerobic hybrid/aerobic CSTR systems.

PubMed

Ağdağ, Osman Nuri

2011-01-01

Leachate generated in municipal solid waste landfill contains large amounts of organic and inorganic contaminants. In the scope of the study, characterization and anaerobic/aerobic treatability of leachate from Denizli (Turkey) Sanitary Landfill were investigated. Time-based fluctuations in characteristics of leachate were monitored during a one-year period. In characterization study; chemical oxygen demand (COD), biochemical oxygen demand (BOD) dissolved oxygen, temperature, pH, alkalinity, volatile fatty acids, total nitrogen, NH4-N, BOD5/COD ratio, suspended solid, inert COD, anaerobic toxicity assay and heavy metals concentrations in leachate were monitored. Average COD, BOD and NH4-N concentration in leachate were measured as 18034 mg/l, 11504 mg/l and 454 mg/l, respectively. Generally, pollution parameters in leachate were higher in summer and relatively lower in winter due to dilution by precipitation. For treatment of leachate, two different reactors, namely anaerobic hybrid and aerobic completely stirred tank reactor (CSTR) having effective volumes of 17.7 and 10.5 litres, respectively, were used. After 41 days of start-up period, leachate was loaded to hybrid reactor at 10 different organic loading rates (OLRs). OLR was increased by increasing COD concentrations. COD removal efficiency of hybrid reactor was carried out at a maximum of 91%. A percentage of 96% of residual COD was removed in the aerobic reactor. NH4-N removal rate in CSTR was quite high. In addition, high methane content was obtained as 64% in the hybrid reactor. At the end of the study, after 170 operation days, it can be said that the hybrid reactor and CSTR were very effective for leachate treatment.

The development of a high sensitivity neutron displacement damage sensor

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

Tonigan, Andrew M.; Parma, Edward J.; Martin, William J.

Here, the capability to characterize the neutron energy spectrum and fluence received by a test object is crucial to under-standing the damage effects observed in electronic components. For nuclear research reactors and high energy density physics fa-cilities this can pose exceptional challenges, especially with low level neutron fluences. An ASTM test method for characterizing neutron environments utilizes the 2N2222A transistor as a 1-MeV equivalent neutron fluence sensor and is applicable for environ-ments with 1 x 10 12 - 1 x 10 14 1-MeV(Si)-Eqv.-n/cm 2. In this work we seek to extend the range of this test method to lower fluencemore » environments utilizing the 2N1486 transistor. Here, the 2N1486 is shown to be an effective neutron displacement damage sensor as low as 1 x 10 10 1-MeV(Si)-Eqv.-n/cm 2.« less

The development of a high sensitivity neutron displacement damage sensor

DOE PAGES

Tonigan, Andrew M.; Parma, Edward J.; Martin, William J.

2016-11-23

Here, the capability to characterize the neutron energy spectrum and fluence received by a test object is crucial to under-standing the damage effects observed in electronic components. For nuclear research reactors and high energy density physics fa-cilities this can pose exceptional challenges, especially with low level neutron fluences. An ASTM test method for characterizing neutron environments utilizes the 2N2222A transistor as a 1-MeV equivalent neutron fluence sensor and is applicable for environ-ments with 1 x 10 12 - 1 x 10 14 1-MeV(Si)-Eqv.-n/cm 2. In this work we seek to extend the range of this test method to lower fluencemore » environments utilizing the 2N1486 transistor. Here, the 2N1486 is shown to be an effective neutron displacement damage sensor as low as 1 x 10 10 1-MeV(Si)-Eqv.-n/cm 2.« less

Massively parallel polymerase cloning and genome sequencing of single cells using nanoliter microwells

PubMed Central

Gole, Jeff; Gore, Athurva; Richards, Andrew; Chiu, Yu-Jui; Fung, Ho-Lim; Bushman, Diane; Chiang, Hsin-I; Chun, Jerold; Lo, Yu-Hwa; Zhang, Kun

2013-01-01

Genome sequencing of single cells has a variety of applications, including characterizing difficult-to-culture microorganisms and identifying somatic mutations in single cells from mammalian tissues. A major hurdle in this process is the bias in amplifying the genetic material from a single cell, a procedure known as polymerase cloning. Here we describe the microwell displacement amplification system (MIDAS), a massively parallel polymerase cloning method in which single cells are randomly distributed into hundreds to thousands of nanoliter wells and simultaneously amplified for shotgun sequencing. MIDAS reduces amplification bias because polymerase cloning occurs in physically separated nanoliter-scale reactors, facilitating the de novo assembly of near-complete microbial genomes from single E. coli cells. In addition, MIDAS allowed us to detect single-copy number changes in primary human adult neurons at 1–2 Mb resolution. MIDAS will further the characterization of genomic diversity in many heterogeneous cell populations. PMID:24213699

Hanford Laboratories Operation monthly activities report, August 1959

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

Not Available

1959-09-15

This is the monthly report for the Hanford Laboratories Operation, August, 1959. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology financial activities, visits, biology operation, physics and instrumentation research, employee relations, and operations research and synthesis operation are discussed.

Hanford Laboratories Operation monthly activities report, September 1961

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

Not Available

1961-10-16

This is the monthly report for the Hanford Laboratories Operation September 1961. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, physics and instrumentation research, operations research and synthesis, programming, and radiation protection operation are discussed.

VIEW OF 77710A REACTOR WING, LOOKING NORTH, SHOWING DOOR TO ...

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

VIEW OF 777-10A REACTOR WING, LOOKING NORTH, SHOWING DOOR TO PROCESS DEVELOPMENT PILE ROOM AND LABORATORY WING ON RIGHT IN BACKGROUND - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

SHIPPINGPORT OPERATIONS FROM POWER OPERATION AFTER FIRST REFUELING TO SECOND REFUELING, MAY 6, 1960 TO AUGUST 16, 1961

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

None

1963-10-31

A report of Shippingport operation during Seed 2 lifetime is presented. The information is primarily confined to the nuclear portion of the operation. A general review of station performance is given along with details of reactor physics, reactor thermal and hydraulic performance, reactor plant performance and modifications, operational chemistry, and radioactive contamination experience. (J.R.D.)

Analysis of granular flow in a pebble-bed nuclear reactor.

PubMed

Rycroft, Chris H; Grest, Gary S; Landry, James W; Bazant, Martin Z

2006-08-01

Pebble-bed nuclear reactor technology, which is currently being revived around the world, raises fundamental questions about dense granular flow in silos. A typical reactor core is composed of graphite fuel pebbles, which drain very slowly in a continuous refueling process. Pebble flow is poorly understood and not easily accessible to experiments, and yet it has a major impact on reactor physics. To address this problem, we perform full-scale, discrete-element simulations in realistic geometries, with up to 440,000 frictional, viscoelastic 6-cm-diam spheres draining in a cylindrical vessel of diameter 3.5m and height 10 m with bottom funnels angled at 30 degrees or 60 degrees. We also simulate a bidisperse core with a dynamic central column of smaller graphite moderator pebbles and show that little mixing occurs down to a 1:2 diameter ratio. We analyze the mean velocity, diffusion and mixing, local ordering and porosity (from Voronoi volumes), the residence-time distribution, and the effects of wall friction and discuss implications for reactor design and the basic physics of granular flow.

D-He-3 spherical torus fusion reactor system study

NASA Astrophysics Data System (ADS)

Macon, William A., Jr.

1992-04-01

This system study extrapolates present physics knowledge and technology to predict the anticipated characteristics of D-He3 spherical torus fusion reactors and their sensitivity to uncertainties in important parameters. Reference cases for steady-state 1000 MWe reactors operating in H-mode in both the 1st stability regime and the 2nd stability regime were developed and assessed quantitatively. These devices would a very small aspect ratio (A=1,2), a major radius of about 2.0 m, an on-axis magnetic field less than 2 T, a large plasma current (80-120 MA) dominated by the bootstrap effect, and high plasma beta (greater than O.6). The estimated cost of electricity is in the range of 60-90 mills/kW-hr, assuming the use of a direct energy conversion system. The inherent safety and environmental advantages of D-He3 fusion indicate that this reactor concept could be competitive with advanced fission breeder reactors and large-scale solar electric plants by the end of the 21st century if research and development can produce the anticipated physics and technology advances.

Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

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

Hess, J. D.; Gauld, I. C.; Gulliford, J.

2017-01-01

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Projectmore » (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.« less

Summary of ORSphere Critical and Reactor Physics Measurements

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

Marshall, Margaret A.; Bess, John D.

In the early 1970s Dr. John T. Mihalczo (team leader), J. J. Lynn, and J. R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVAmore » I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.« less

Experimental Anomalies in Neutrino Physics

NASA Astrophysics Data System (ADS)

Palamara, Ornella

2014-03-01

In recent years, experimental anomalies ranging in significance (2.8-3.8 σ) have been reported from a variety of experiments studying neutrinos over baselines less than 1 km. Results from the LSND and MiniBooNE short-baseline νe /νe appearance experiments show anomalies which cannot be described by oscillations between the three standard model neutrinos (the ``LSND anomaly''). In addition, a re-analysis of the anti-neutrino flux produced by nuclear power reactors has led to an apparent deficit in νe event rates in a number of reactor experiments (the ``reactor anomaly''). Similarly, calibration runs using 51Cr and 37Ar radioactive sources in the Gallium solar neutrino experiments GALLEX and SAGE have shown an unexplained deficit in the electron neutrino event rate over very short distances (the ``Gallium anomaly''). The puzzling results from these experiments, which together may suggest the existence of physics beyond the Standard Model and hint at exciting new physics, including the possibility of additional low-mass sterile neutrino states, have raised the interest in the community for new experimental efforts that could eventually solve this puzzle. Definitive evidence for sterile neutrinos would be a revolutionary discovery, with implications for particle physics as well as cosmology. Proposals to address these signals by employing accelerator, reactor and radioactive source experiments are in the planning stages or underway worldwide. In this talk some of these will be reviewed, with emphasis on the accelerator programs.

Automated Characterization of Rotating MHD Modes and Subsequent Locking in a Tokamak

NASA Astrophysics Data System (ADS)

Riquezes, Juan; Sabbagh, Steven; Berkery, Jack

2016-10-01

Disruption avoidance in tokamaks is highly desired to maintain steady plasma operation, and is critical for future reactor-scale devices, such as ITER, to avoid potential damage to device components. This high priority research is being conducted at PPPL by analyzing data from NSTX and its upgrade, NSTX-U. A key cause of disruptions is the physical event chain that comprises the appearance of rotating MHD modes, their slowing by resonant field drag mechanisms, and their subsequent locking. The present research aims to define algorithms to automatically find and characterize such physical event chains in the machine database. Characteristics such as identification of a mode locking time based on a loss of torque balance and bifurcation of the mode rotation frequency are examined to determine the reliability of such events in predicting disruptions. A goal is to detect such behavior as early as possible during a plasma discharge, and to further examine potential ways to forecast it. This capability could be used to provide a warning to use active mode control as a disruption avoidance mechanism, or to trigger a controlled plasma shutdown if desired. Supported by US DOE Contracts DE-FG02-99ER54524 and DE-AC02-09CH11466.

Wood liquefaction and its application to Novolac resin

Treesearch

Hui Pan; Chung-Yun Hse; Todd F. Shupe

2009-01-01

Wood liquefaction was conducted using phenol as a reagent solvent with a weak acid catalyst in two different reactors: (Alma et al., 1995a.) an atmospheric glass reactor and (Alma et al., 1995b.) a sealed Parr® reactor. Residues were characterized by wet chemical analyses, Fourier transform infrared (FT-IR) spectroscopy, and X-ray diffraction (XRD). The FT-IR...

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.

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

Smith, Curtis L.; Prescott, Steven; Coleman, Justin

This report describes the current progress and status related to the Industry Application #2 focusing on External Hazards. For this industry application within the Light Water Reactor Sustainability (LWRS) Program Risk-Informed Safety Margin Characterization (RISMC) R&D Pathway, we will create the Risk-Informed Margin Management (RIMM) approach to represent meaningful (i.e., realistic facility representation) event scenarios and consequences by using an advanced 3D facility representation that will evaluate external hazards such as flooding and earthquakes in order to identify, model and analyze the appropriate physics that needs to be included to determine plant vulnerabilities related to external events; manage the communicationmore » and interactions between different physics modeling and analysis technologies; and develop the computational infrastructure through tools related to plant representation, scenario depiction, and physics prediction. One of the unique aspects of the RISMC approach is how it couples probabilistic approaches (the scenario) with mechanistic phenomena representation (the physics) through simulation. This simulation-based modeling allows decision makers to focus on a variety of safety, performance, or economic metrics. In this report, we describe the evaluation of various physics toolkits related to flooding representation. Ultimately, we will be coupling the flooding representation with other events such as earthquakes in order to provide coupled physics analysis for scenarios where interactions exist.« less

Thermionic fast spectrum reactor-converter on the basis of multi-cell TFE

NASA Astrophysics Data System (ADS)

Ponomarev-Stepnoi, N. N.; Kompaniets, G. V.; Poliakov, D. N.; Stepennov, B. S.; Andreev, P. V.; Zhabotinsky, E. E.; Nikolaev, Yu. V.; Lapochkin, N. V.

2001-02-01

Today Russian experts have technological experience in development of in-core thermionic converters for reactors of space nuclear power plants. Such a converter contains nuclear fuel inside and really represents a fuel element of a reactor. Two types of reactors can be considered on the basis of these thermionic fuel elements: with thermal or intermediate neutron spectrum, and with fast neutron spectrum. The first type is characterized by the presence of moderator in core that ensures most economical usage of nuclear fuel. The estimation shows that moderated system is the most effective in the power range of about 5 ... 100 kWe. The power systems of higher level are characterized by larger dimensions due to the presence of moderator. The second type of reactor is considered for higher power levels. This power range is about hundreds kWe. Dimensions of the fast reactor and core configuration are determined by the necessity to ensure the required net output power, on the one hand, and the necessity to ensure critical state on the other hand. In the case of using in-core thermionic fuel elements of the specified design, minimal reactor output power is determined by reactor criticality condition, and maximum reactor power output is determined by specifications and launcher capabilities. In the present paper the effective multiplication factor of a fast spectrum reactor on the basis of a multi-cell TFE developed by ``Lutch'' is considered a function of the total number of TFEs in the reactor. The MCU Monte-Carlo code, developed in Russia (Alekseev, et al., 1991), was used for computations. TFE computational models are placed in the nodes of a uniform triangular lattice and surrounded with pressure vessel and a side reflector. Ordinary fuel pins without thermionic converters were used instead of some TFEs to optimize criticality parameters, dimensions and output power of the reactor. General weight parameters of the reactor are presented in the paper. .

Applications of nuclear physics

NASA Astrophysics Data System (ADS)

Hayes, A. C.

2017-02-01

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applications of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.

Applications of nuclear physics

DOE PAGES

Hayes-Sterbenz, Anna Catherine

2017-01-10

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applicationsmore » of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Lastly, each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.« less

Applications of nuclear physics.

PubMed

Hayes, A C

2017-02-01

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applications of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.

Applications of nuclear physics

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

Hayes-Sterbenz, Anna Catherine

Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applicationsmore » of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Lastly, each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.« less

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

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

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

Hanford Atomic Products Operation monthly report for March 1956

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

Not Available

1956-04-20

This is the monthly report for the Hanford Laboratories Operation, March, 1956. Metallurgy, reactor fuels, chemistry, dosimetry, separation processes, reactor technology; financial activities, visits, biology operation, physics and instrumentation research, employee relations, pile technology, safety and radiological sciences are discussed.

A Course in Chemical Reactor Design.

ERIC Educational Resources Information Center

Takoudis, Christos G.

1983-01-01

Presents course outline, topics covered, and final project (doubling as a take home final exam) for a one-semester, interdisciplinary course on the design and behavior of chemical reactors. Interplay of chemical and physical rate processes is stressed in the course. (JM)

SCALE Code System

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

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

SCALE Code System 6.2.1

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

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.« less

Development of Subspace-based Hybrid Monte Carlo-Deterministric Algorithms for Reactor Physics Calculations

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

Abdel-Khalik, Hany S.; Zhang, Qiong

2014-05-20

The development of hybrid Monte-Carlo-Deterministic (MC-DT) approaches, taking place over the past few decades, have primarily focused on shielding and detection applications where the analysis requires a small number of responses, i.e. at the detector locations(s). This work further develops a recently introduced global variance reduction approach, denoted by the SUBSPACE approach is designed to allow the use of MC simulation, currently limited to benchmarking calculations, for routine engineering calculations. By way of demonstration, the SUBSPACE approach is applied to assembly level calculations used to generate the few-group homogenized cross-sections. These models are typically expensive and need to be executedmore » in the order of 10 3 - 10 5 times to properly characterize the few-group cross-sections for downstream core-wide calculations. Applicability to k-eigenvalue core-wide models is also demonstrated in this work. Given the favorable results obtained in this work, we believe the applicability of the MC method for reactor analysis calculations could be realized in the near future.« less

Report detailing comparative analysis of results from high flux isotope reactor and national institute of standards technology small-angle neutron scattering experiments

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

Sokolov, Mikhail A.; Littrell, Ken; Wells, Peter

The major issues regarding irradiation effects are discussed in [1-3] and have also been discussed in previous progress and milestone reports. As noted previously, of the many significant issues discussed, the issue considered to have the most impact on the current regulatory process is that associated with effects of neutron irradiation on RPV steels at high fluence, for long irradiation times, and as affected by neutron flux. It is clear that embrittlement of RPV steels is a critical issue that may limit LWR plant life extension. The primary objective of the LWRSP RPV task is to develop robust predictions ofmore » transition temperature shifts (TTS) at high fluence ( t) to at least 1020 n/cm 2 (>1 MeV) pertinent to plant operation of some pressurized water reactors (PWR) for 80 full power years. Correlations between the high flux test reactor results and low flux surveillance specimens must be established for proper RPV embrittlement predictions of the current nuclear power fleet. Additionally, a complete understanding of defect evolution for high nickel RPV steels is needed to characterize the embrittlement potential of Mn-Ni-enriched precipitates (MNPs), particularly for the high fluence regime. While understanding of copper-enriched precipitates (CRPs) have been fully developed, the recent discovery and experimental verification [4] of late blooming MNPs with little to no copper for nucleation has stimulated research efforts to understand the evolution of these phases. New and existing databases will be combined to support developing physically based models of TTS for high fluence-low flux ( < 10 11n/cm 2-s) conditions, beyond the existing surveillance database, to neutron fluences of at least 1 1020 n/cm2 (>1 MeV). Moreover, large number of various RPV materials have been irradiated in ATR-2 experiment and will be jointly studied by University of California Santa Barbara (UCSB) and ORNL to address majority of microstructural characteristics discussed above, see Ref. [5] and [6] for details. UCSB has performed a large number of SANS experiments in the past at the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR). These data are taken from RPV steels irradiated in a wide range of flux-fluence space and will be very useful in comparing to the upcoming UCSB ATR-2 irradiation characterization since most of the SANS experiments with ATR-2 materials will be performed at ORNL High Flux Isotope Reactor (HFIR). However in the previous report [7], some discrepancies were observed between HFIR and NCNR generated data. One of the hypotheses was that there was some kind of extra scattering occurring off the sample holders that results in the HFIR curves falling above the NCNR curves. To test this hypothesis, UCSB provided thermally aged samples that have been previously run at NCNR to ORNL for testing at HFIR while ORNL performed some improvements to experimental set up at HFIR. This report provides the status for the Level 3 Milestone (M3LW-15OR0402013), Complete report detailing comparative analysis of results from High Flux Isotope Reactor and National Institute of Standards and Technology small-angle neutron scattering experiments. This milestone is associated with small-angle neutron scattering characterization at the High Flux Isotope Reactor of various model alloys that had been previously characterized at NCNR by UCSB.« less

Temperature measuring analysis of the nuclear reactor fuel assembly

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

Urban, F., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Kučák, L., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk; Bereznai, J., E-mail: jozef.bereznai@stuba.sk, E-mail: zdenko.zavodny@stuba.sk

2014-08-06

Study was based on rapid changes of measured temperature values from the thermocouple in the VVER 440 nuclear reactor fuel assembly. Task was to determine origin of fluctuations of the temperature values by experiments on physical model of the fuel assembly. During an experiment, heated water was circulating in the system and cold water inlet through central tube to record sensitivity of the temperature sensor. Two positions of the sensor was used. First, just above the central tube in the physical model fuel assembly axis and second at the position of the thermocouple in the VVER 440 nuclear reactor fuelmore » assembly. Dependency of the temperature values on time are presented in the diagram form in the paper.« less

REACTOR

DOEpatents

Christy, R.F.

1961-07-25

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

Babcock and Wilcox assessment of the Pratt and Whitney XNR2000

NASA Technical Reports Server (NTRS)

Westerman, Kurt O.; Scoles, Stephen W.; Jensen, R. R.; Rodes, J. R.; Ales, M. W.

1993-01-01

Babcock & Wilcox performed four subtasks related to the assessment of the Pratt & Whitney XNR2000 nuclear reactor as follows: (1) cermet fuel element fabricability assessment; (2) mechanical design review of the reactor system; (3) neutronic analysis review; and (4) safety assessment. The results of the mechanical and physics reviews have been integrated into the reactor design. The results of the fuel and safety assessments are presented.

Characteristics of aerobic granules grown on glucose a sequential batch shaking reactor.

PubMed

Cai, Chun-guang; Zhu, Nan-wen; Liu, Jun-shen; Wang, Zhen-peng; Cai, Wei-min

2004-01-01

Aerobic heterotrophic granular sludge was cultivated in a sequencing batch shaking reactor (SBSR) in which a synthetic wastewater containing glucose as carbon source was fed. The characteristics of the aerobic granules were investigated. Compared with the conventional activated sludge flocs, the aerobic granules exhibit excellent physical characteristics in terms of settleability, size, shape, biomass density, and physical strength. Scanning electron micrographs revealed that in mature granules little filamentous bacteria could be found, rod-shaped and coccoid bacteria were the dominant microorganisms.

VIEW OF STEEL PLATE DOOR IN NUCLEAR PHYSICS LABORATORY, BETWEEN ...

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

VIEW OF STEEL PLATE DOOR IN NUCLEAR PHYSICS LABORATORY, BETWEEN LABORATORY AND SP-SE REACTOR ROOM,LEVEL -15’, LOOKING NORTHWEST - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

Brookhaven highlights, October 1978-September 1979. [October 1978 to September 1979

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

Not Available

1979-01-01

These highlights present an overview of the major research and development achievements at Brookhaven National Laboratory from October 1978 to September 1979. Specific areas covered include: accelerator and high energy physics programs; high energy physics research; the AGS and improvements to the AGS; neutral beam development; heavy ion fusion; superconducting power cables; ISABELLE storage rings; the BNL Tandem accelerator; heavy ion experiments at the Tandem; the High Flux Beam Reactor; medium energy physics; nuclear theory; atomic and applied physics; solid state physics; neutron scattering studies; x-ray scattering studies; solid state theory; defects and disorder in solids; surface physics; the Nationalmore » Synchrotron Light Source ; Chemistry Department; Biology Department; Medical Department; energy sciences; environmental sciences; energy technology programs; National Center for Analysis of Energy Systems; advanced reactor systems; nuclear safety; National Nuclear Data Center; nuclear materials safeguards; Applied Mathematics Department; and support activities. (GHT)« less

Physical models and primary design of reactor based slow positron source at CMRR

NASA Astrophysics Data System (ADS)

Wang, Guanbo; Li, Rundong; Qian, Dazhi; Yang, Xin

2018-07-01

Slow positron facilities are widely used in material science. A high intensity slow positron source is now at the design stage based on the China Mianyang Research Reactor (CMRR). This paper describes the physical models and our primary design. We use different computer programs or mathematical formula to simulate different physical process, and validate them by proper experiments. Considering the feasibility, we propose a primary design, containing a cadmium shield, a honeycomb arranged W tubes assembly, electrical lenses, and a solenoid. It is planned to be vertically inserted in the Si-doping channel. And the beam intensity is expected to be 5 ×109

78 FR 59981 - Proposed Revision to Physical Security-Standard Design Certification and Operating Reactors

Federal Register 2010, 2011, 2012, 2013, 2014

2013-09-30

... the Standard Review Plan (SRP), concerning the physical security reviews of design certification... NRC staff with the physical security review of applications for design certifications, incorporate... NUCLEAR REGULATORY COMMISSION [NRC-2013-0225] Proposed Revision to Physical Security--Standard...

VIEW OF 77710A REACTOR WING, LOOKING NORTHEAST,SHOWING LOADING DOOR TO ...

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

VIEW OF 777-10A REACTOR WING, LOOKING NORTHEAST,SHOWING LOADING DOOR TO THE PROCESS DEVELOPMENT PILE ROOM. BUILDING 305-A IN BACKGROUND ON LEFT - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

Semiconductor Chemical Reactor Engineering and Photovoltaic Unit Operations.

ERIC Educational Resources Information Center

Russell, T. W. F.

1985-01-01

Discusses the nature of semiconductor chemical reactor engineering, illustrating the application of this engineering with research in physical vapor deposition of cadmium sulfide at both the laboratory and unit operations scale and chemical vapor deposition of amorphous silicon at the laboratory scale. (JN)

Site remediation techniques in India: a review

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

Anomitra Banerjee; Miller Jothi

India is one of the developing countries operating site remediation techniques for the entire nuclear fuel cycle waste for the last three decades. In this paper we intend to provide an overview of remediation methods currently utilized at various hazardous waste sites in India, their advantages and disadvantages. Over the years the site remediation techniques have been well characterized and different processes for treatment, conditioning and disposal are being practiced. Remediation Methods categorized as biological, chemical or physical are summarized for contaminated soils and environmental waters. This paper covers the site remediation techniques implemented for treatment and conditioning of wastelandsmore » arising from the operation of nuclear power plant, research reactors and fuel reprocessing units. (authors)« 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.

Investigation report: H Reactor mischarging incident

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

Vinther, A.P.

1964-05-01

All cold reactor start-up procedures require the vertical safety rods (VSR) to be withdrawn in pairs with specific waiting periods between each pair withdrawal. This rod withdrawal procedure will assure an early and safe detection of reactor criticality should reactor reactivity conditions be different than predicted so that proper corrective actions can be taken. During the paired VSR removal of H reactor on April 17, 1964, while preparing for reactor start-up, an extremely low level rising period was detected with six VSR's still in the reactor. The withdrawn VSR's were promptly re-inserted. During the next several days other process difficultiesmore » were encountered. H Processing personnel began investigating the possibility that a number of process tubes might have been mischarged; one shift's charging effort appeared to be suspect as longitudinal peaking appeared nearly twice as severe as normal in the distorted region. Following verification of a charging error in the suspect group of 171 tubes, that group of tubes was discharged and recharged with the proper charge make-up. On April 24, during VSR removal for start-up, low level criticality was detected with three VSR's still in the unit. The VSR's were re-inserted and Operational Physics analysis requested. Following installation of additional poisoning, the Operational Physics analysis uncovered a reactivity prediction error related to the prior operation with the skewed flux distribution. However, in this case, as on April 17, the procedural paired VSR withdrawal provided safe detection of the criticality condition in adequate time to take prompt corrective action. A successful reactor start-up was then achieved later on April 24, and reactor operation has been normal since that time. 4 figs.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

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

Mertyurek, Ugur; Gauld, Ian C.

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Development of ORIGEN Libraries for Mixed Oxide (MOX) Fuel Assembly Designs

DOE PAGES

Mertyurek, Ugur; Gauld, Ian C.

2015-12-24

In this research, ORIGEN cross section libraries for reactor-grade mixed oxide (MOX) fuel assembly designs have been developed to provide fast and accurate depletion calculations to predict nuclide inventories, radiation sources and thermal decay heat information needed in safety evaluations and safeguards verification measurements of spent nuclear fuel. These ORIGEN libraries are generated using two-dimensional lattice physics assembly models that include enrichment zoning and cross section data based on ENDF/B-VII.0 evaluations. Using the SCALE depletion sequence, burnup-dependent cross sections are created for selected commercial reactor assembly designs and a representative range of reactor operating conditions, fuel enrichments, and fuel burnup.more » The burnup dependent cross sections are then interpolated to provide problem-dependent cross sections for ORIGEN, avoiding the need for time-consuming lattice physics calculations. The ORIGEN libraries for MOX assembly designs are validated against destructive radiochemical assay measurements of MOX fuel from the MALIBU international experimental program. This program included measurements of MOX fuel from a 15 × 15 pressurized water reactor assembly and a 9 × 9 boiling water reactor assembly. The ORIGEN MOX libraries are also compared against detailed assembly calculations from the Phase IV-B numerical MOX fuel burnup credit benchmark coordinated by the Nuclear Energy Agency within the Organization for Economic Cooperation and Development. Finally, the nuclide compositions calculated by ORIGEN using the MOX libraries are shown to be in good agreement with other physics codes and with experimental data.« less

Response to "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

2014-05-01

Relying on coil positions relative to the plasma, the "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the "proximity condition," used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

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

Belles, Randy; Jain, Prashant K.; Powers, Jeffrey J.

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments usingmore » equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.« less

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.

Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures.

PubMed

Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R; Crowhurst, Jonathan C; Weisz, David G; Zaug, Joseph M; Dai, Zurong; Radousky, Harry B; Chernov, Alex; Ramon, Erick; Stavrou, Elissaios; Knight, Kim; Fabris, Andrea L; Cappelli, Mark A; Rose, Timothy P

2017-09-01

We present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after they pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.

Recent upgrades and new scientific infrastructure of MARIA research reactor, Otwock-Swierk, Poland

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

NONE

The MARIA reactor is open-pool type, water and beryllium moderated. It has two independent primary cooling systems: fuel and pool cooling system. Each fuel assembly is cooled down separately in pressurized channels with individual performances characterization. The fuel assemblies consist of five layers of bent plates or six concentric tubes. Currently it is one of the most powerful research reactors in Europe with operation availability at least up to 2030. Its nominal thermal power is 30 MW. It is characterized by high neutron flux density: up to 3x10{sup 14} n cm{sup -2} s{sup -1} in case of thermal neutrons, andmore » up to 2x10{sup 13} n cm{sup -2} s{sup -1} in case of fast neutrons. The reactor is operated for ca. 4000 h per year. The reactor facility is equipped with fully equipped three hot cells with shielding up to 10{sup 15} Bq. Adjacent to the reactor facility, the radio-pharmaceutics plant (POLATOM) and Material Research Laboratory are located. They are equipped with a number of hot cells with instrumentation. The transport system of radioactive materials from reactor facility to Material Research Laboratory is available. During 2014 the MARIA reactor has been operated with three different types of fuel the same time: previous 36% enriched fuel, and two types of new LEU fuels. In the meantime, molybdenum irradiation programme has been developed. Maria is a multifunctional research tool, with a notable application in production of radioisotopes, radio-pharmaceutics manufacturing (ca. 600 TBq/y), {sup 99}Mo for medical scintigraphy (ca. 6000 TBq/y), neutron transmutation doping of silicon single crystals, wide scientific research based on neutron beams utilization. From the beginning MARIA reactor was intended for loop and fuel testing research activities. Currently it is used mostly as material testing and irradiation facility and for that reason it has wide experimental capabilities. There are eight horizontal irradiation channels from among whom six of them are equipped with instrumentation for condensed matter physics research: - H3 - spectrometer and diffractometer with double monochromator; - H4 - small angle scattering spectrometer; - H5 - polarized neutrons spectrometer; - H6, H7 - two 3-axial crystal neutron spectrometers; - H8 - neutron radiography stand. For two horizontal channels are ongoing exploitation programs: - H2 - station with epithermal neutron beam produced in uranium converter is being developed. Intelligent converter will be installed on the periphery of reactor core. The intensity of the beam will be at the level 2x10{sup 9} n cm{sup -2}s{sup -1} what makes the beam unique in the Europe. - H1 - special pneumatic horizontal mail is being developed for irradiation material samples in the vicinity of the core i.e. in the distal part of the H1 channel. The number of neutron irradiation facilities in MARIA reactor is increasing every year. Numerous of thermal neutron irradiation channels including fast hydraulic rabbit system and large size channels for fast neutron irradiation are used routinely. Recently new in-pile facility with ITER-like neutron energy spectrum for 14 MeV neutron irradiation has been constructed. Taking into account its performance and ability of almost incessant operation the facility appears as one of the most powerful 14 MeV neutron sources. The facility shall be used for material research connected with thermonuclear devices (ITER) and 4. generation nuclear reactors. The system of independent fuels channels used in MARIA reactor appear to be very flexible and very convenient to be used as irradiation channels for uranium targets for {sup 99}Mo production. Currently, MARIA reactor supplies ca. 18% world production of {sup 99}Mo. The MARIA reactor research activities are still extended. The current scientific projects are connected e.g. with silicon neutron transmutation doping, in-pile gamma heating measurements, French calculation codes implementation (TRIPOLI4, APOLLO2). The horizontal neutron beams utilization is also developed. The MARIA reactor, due to its primary application connected with loop and fuel testing, is very convenient for testing the nuclear instrumentation, control and measurement systems.« less

Equipment for neutron measurements at VR-1 Sparrow training reactor.

PubMed

Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

2010-01-01

The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

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

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

D.M. McEligot; K. G. Condie; G. E. McCreery

2005-10-01

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

Nuclear reactor target assemblies, nuclear reactor configurations, and methods for producing isotopes, modifying materials within target material, and/or characterizing material within a target material

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

Toth, James J.; Wall, Donald; Wittman, Richard S.

Target assemblies are provided that can include a uranium-comprising annulus. The assemblies can include target material consisting essentially of non-uranium material within the volume of the annulus. Reactors are disclosed that can include one or more discrete zones configured to receive target material. At least one uranium-comprising annulus can be within one or more of the zones. Methods for producing isotopes within target material are also disclosed, with the methods including providing neutrons to target material within a uranium-comprising annulus. Methods for modifying materials within target material are disclosed as well as are methods for characterizing material within a targetmore » material.« less

Neutron flux and power in RTP core-15

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

Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Zin, Muhammad Rawi Md; Usang, Mark Dennis

PUSPATI TRIGA Reactor achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution of TRIGA core. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core withmore » literally no physical approximation. The consistency and accuracy of the developed RTP MCNP model was established by comparing calculations to the available experimental results and TRIGLAV code calculation.« less

Flow Reactor for studying Physicochemical and aging properties of SOA

NASA Astrophysics Data System (ADS)

Babar, Z. B.

2016-12-01

Secondary organic aerosols (SOA) have importance in environmental processes such as affecting earth's radiative balance and cloud formation processes. For studying SOA formation large scale environmental batch reactors and laboratory scale flow reactors have been used. In this study application of flow reactor to study physicochemical properties of SOA is also investigated after its characterization. The flow reactor is of cylindrical design (ID 15 cm x L 70 cm) equipped with UV lamps. It is coupled with various instruments such as scanning mobility particle sizer, NOx analyzer, ozone analyzer, VOC analyzer, hygrometer, and temperature sensors for gas and particle phase measurements. OH radicals were generated by custom build ozone generator and relative humidity. The following characterizations were performed: (1) residence time distribution (RTD) measurements, (2) RH and temperature control, (3) OH radical exposure range (atmospheric aging time), (4) gas phase oxidation of SOA precursors such as α-pinene by OH radical. The flow reactor yielded narrow RTDs. In particular, RH and temperature can be controlled effectively between 0-60% and 22-43oC, respectively. OH radical exposure ranges from 6.49x1010 to 3.68x1011 molecules/cm3s (0.49 to 4.91 days). Our initial efforts on OH radical generation using hydrogen peroxide and its quantification by using flourescenet technique will be also be presented.

Rebuilding the Brookhaven high flux beam reactor: A feasibility study

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

Brynda, W.J.; Passell, L.; Rorer, D.C.

1995-01-01

After nearly thirty years of operation, Brookhaven`s High Flux Beam Reactor (HFBR) is still one of the world`s premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR`s value as a nationalmore » scientific resource, members of the Laboratory`s scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor`s research capabilities.« less

Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

NASA Astrophysics Data System (ADS)

Kulikov, E. G.

2017-01-01

Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

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

Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.

This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {supmore » 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.« less

10 CFR 72.180 - Physical protection plan.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

10 CFR 72.180 - Physical protection plan.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

10 CFR 72.180 - Physical protection plan.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

10 CFR 72.180 - Physical protection plan.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

10 CFR 72.180 - Physical protection plan.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Physical protection plan. 72.180 Section 72.180 Energy... NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE, AND REACTOR-RELATED GREATER THAN CLASS C WASTE Physical Protection § 72.180 Physical protection plan. The licensee shall establish, maintain, and follow a detailed...

Current Physics Research: Part I.

ERIC Educational Resources Information Center

Schewe, Phillip F.

1980-01-01

This article is a preview of the book, "Physics News in 1980." Five research areas are reviewed: high energy particle accelerators, fusion reactors, solar cells, astrophysics, and gauge theories. (Author/DS)

MONTE CARLO SIMULATIONS OF PERIODIC PULSED REACTOR WITH MOVING GEOMETRY PARTS

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

Cao, Yan; Gohar, Yousry

2015-11-01

In a periodic pulsed reactor, the reactor state varies periodically from slightly subcritical to slightly prompt supercritical for producing periodic power pulses. Such periodic state change is accomplished by a periodic movement of specific reactor parts, such as control rods or reflector sections. The analysis of such reactor is difficult to perform with the current reactor physics computer programs. Based on past experience, the utilization of the point kinetics approximations gives considerable errors in predicting the magnitude and the shape of the power pulse if the reactor has significantly different neutron life times in different zones. To accurately simulate themore » dynamics of this type of reactor, a Monte Carlo procedure using the transfer function TRCL/TR of the MCNP/MCNPX computer programs is utilized to model the movable reactor parts. In this paper, two algorithms simulating the geometry part movements during a neutron history tracking have been developed. Several test cases have been developed to evaluate these procedures. The numerical test cases have shown that the developed algorithms can be utilized to simulate the reactor dynamics with movable geometry parts.« less

Conceptual Design and Neutronics Analyses of a Fusion Reactor Blanket Simulation Facility

DTIC Science & Technology

1986-01-01

Laboratory (LLL) ORNL Oak Ridge National Laboratory PPPL Princeton Plasma Physics Laboratory RSIC Reactor Shielding Information Center (at ORNL) SS...Module (LBM) to be placed in the TFTR at PPPL . Jassby et al. describe the program, including design, manufacturing techniques. neutronics analyses, and

Cryogenic Hydrogen Fuel for Controlled Inertial Confinement Fusion (Cryogenic Target Factory Concept Based on FST-Layering Method)

NASA Astrophysics Data System (ADS)

Aleksandrova, I. V.; Koresheva, E. R.; Koshelev, I. E.; Krokhin, O. N.; Nikitenko, A. I.; Osipov, I. E.

2017-12-01

A central element of a power plant based on inertial confinement fusion (ICF) is a target with cryogenic hydrogen fuel that should be delivered to the center of a reactor chamber with a high accuracy and repetition rate. Therefore, a cryogenic target factory (CTF) is an integral part of any ICF reactor. A promising way to solve this problem consists in the FST layering method developed at the Lebedev Physical Institute (LPI). This method (rapid fuel layering inside moving free-standing targets) is unique, having no analogs in the world. The further development of FST-layering technologies is implemented in the scope of the LPI program for the creation of a modular CTF and commercialization of the obtained results. In this report, we discuss our concept of CTF (CTF-LPI) that exhibits the following distinctive features: using a FST-layering technology for the elaboration of an in-line production of cryogenic targets, using an effect of quantum levitation of high-temperature superconductors (HTSCs) in magnetic field for noncontacting manipulation, transport, and positioning of the free-standing cryogenic targets, as well as in using a Fourier holography technique for an on-line characterization and tracking of the targets flying into the reactor chamber. The results of original experimental and theoretical investigations performed at LPI indicate that the existing and developing target fabrication capabilities and technologies can be applied to ICF target production. The unique scientific, engineering, and technological base developed in Russia at LPI allows one to make a CTFLPI prototype for mass production of targets and delivery thereof at the required velocity into the ICF reactor chamber.

Isotopic composition and neutronics of the Okelobondo natural reactor

NASA Astrophysics Data System (ADS)

Palenik, Christopher Samuel

The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

VERAIn

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

Simunovic, Srdjan

2015-02-16

CASL's modeling and simulation technology, the Virtual Environment for Reactor Applications (VERA), incorporates coupled physics and science-based models, state-of-the-art numerical methods, modern computational science, integrated uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs), single-effect experiments, and integral tests. The computational simulation component of VERA is the VERA Core Simulator (VERA-CS). The core simulator is the specific collection of multi-physics computer codes used to model and deplete a LWR core over multiple cycles. The core simulator has a single common input file that drives all of the different physics codes. The parser code, VERAIn, converts VERAmore » Input into an XML file that is used as input to different VERA codes.« less

A complete dosimetry experimental program in support to the core characterization and to the power calibration of the CABRI reactor. A complete dosimetry experimental program in support of the core characterization and of the power calibration of the CABRI reactor

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

Rodiac, F.; Hudelot, JP.; Lecerf, J.

CABRI is an experimental pulse reactor operated by CEA at the Cadarache research center. Since 1978 the experimental programs have aimed at studying the fuel behavior under Reactivity Initiated Accident (RIA) conditions. Since 2003, it has been refurbished in order to be able to provide RIA and LOCA (Loss Of Coolant Accident) experiments in prototypical PWR conditions (155 bar, 300 deg. C). This project is part of a broader scope including an overall facility refurbishment and a safety review. The global modification is conducted by the CEA project team. It is funded by IRSN, which is conducting the CIP experimentalmore » program, in the framework of the OECD/NEA project CIP. It is financed in the framework of an international collaboration. During the reactor restart, commissioning tests are realized for all equipment, systems and circuits of the reactor. In particular neutronics and power commissioning tests will be performed respectively in 2015 and 2016. This paper focuses on the design of a complete and original dosimetry program that was built in support to the CABRI core characterization and to the power calibration. Each one of the above experimental goals will be fully described, as well as the target uncertainties and the forecasted experimental techniques and data treatment. (authors)« less

Computer modeling and simulators as part of university training for NPP operating personnel

NASA Astrophysics Data System (ADS)

Volman, M.

2017-01-01

This paper considers aspects of a program for training future nuclear power plant personnel developed by the NPP Department of Ivanovo State Power Engineering University. Computer modeling is used for numerical experiments on the kinetics of nuclear reactors in Mathcad. Simulation modeling is carried out on the computer and full-scale simulator of water-cooled power reactor for the simulation of neutron-physical reactor measurements and the start-up - shutdown process.

Analysis of the uncertainties in the physical calculations of water-moderated power reactors of the VVER type by the parameters of models of preparing few-group constants

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

Bryukhin, V. V., E-mail: bryuhin@yandex.ru; Kurakin, K. Yu.; Uvakin, M. A.

The article covers the uncertainty analysis of the physical calculations of the VVER reactor core for different meshes of the reference values of the feedback parameters (FBP). Various numbers of nodes of the parametric axes of FBPs and different ranges between them are investigated. The uncertainties of the dynamic calculations are analyzed using RTS RCCA ejection as an example within the framework of the model with the boundary conditions at the core inlet and outlet.

Reactor physics teaching and research in the Swiss nuclear engineering master

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

Chawla, R.; Paul Scherrer Inst., CH-5232 Villigen PSI

Since 2008, a Master of Science program in Nuclear Engineering (NE) has been running in Switzerland, thanks to the combined efforts of the country's key players in nuclear teaching and research, viz. the Swiss Federal Inst.s of Technology at Lausanne (EPFL) and at Zurich (ETHZ), the Paul Scherrer Inst. (PSI) at Villigen and the Swiss Nuclear Utilities (Swissnuclear). The present paper, while outlining the academic program as a whole, lays emphasis on the reactor physics teaching and research training accorded to the students in the framework of the developed curriculum. (authors)

Reactor monitoring with Neutrinos

NASA Astrophysics Data System (ADS)

Cribier, Michel

2011-12-01

The fundamental knowledge on neutrinos acquired in the recent years open the possibility of applied neutrino physics. Among it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable to IAEA in charge of the control of nuclear power plants. Several efforts worldwide have already started.

VIEW OF PROCESS DEVELOPMENT PILE (PDP) TANK, LOOKING WESTSOUTHWEST, BASEMENT ...

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

VIEW OF PROCESS DEVELOPMENT PILE (PDP) TANK, LOOKING WEST-SOUTHWEST, BASEMENT LEVEL -15’. EDGE O FRESONANCE TEST REACTOR (RTR), LATER KNOWN AS LATTICE TEST REACTOR (LTR), VISIBLE TO RIGHT OF PDP TANK - Physics Assembly Laboratory, Area A/M, Savannah River Site, Aiken, Aiken County, SC

Neutronic reactor

DOEpatents

Carleton, John T.

1977-01-25

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

CALIBRATION OF FULL-SCALE OZONATION SYSTEMS WITH CONSERVATIVE AND REACTIVE TRACERS

EPA Science Inventory

A full-scale ozonation reactor was characterized with respect to the overall oxidation budget by coupling laboratory kinetics with reactor hydraulics. The ozone decomposition kinetics and the ratio of the OH radical to the ozone concentration were determined in laboratory batch ...

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

Not Available

The SPS Concept Development and Evaluation Program includes a comparative assessment. An early first step in the assessment process is the selection and characterization of alternative technologies. This document describes the cost and performance (i.e., technical and environmental) characteristics of six central station energy alternatives: (1) conventional coal-fired powerplant; (2) conventional light water reactor (LWR); (3) combined cycle powerplant with low-Btu gasifiers; (4) liquid metal fast breeder reactor (LMFBR); (5) photovoltaic system without storage; and (6) fusion reactor.

Experimental evidences of 95 mTc production in a nuclear reactor.

PubMed

Cohen, I M; Robles, A; Mendoza, P; Airas, R M; Montoya, E H

2018-05-01

95 m Tc has been identified as by-product in some solutions of 99 m Tc obtained by irradiation of molybdenum trioxide in a reactor neutron flux. The characterization was carried out using both measurements by gamma spectrometry and half-life determination. The possible ways that lead to the 95 m Tc production in a nuclear reactor are discussed. Copyright © 2018. Published by Elsevier Ltd.

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.

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

Little, G.A.

For better than ten years there was little public notice of the TRIGA reactor at UC-Berkeley. Then: a) A non-student persuaded the Student and Senate to pass a resolution to request Campus Administration to stop operation of the reactor and remove it from campus. b) Presence of the reactor became a campaign-issue in a City Mayoral election. c) Two local residents reported adverse physical reactions before, during, and after a routine tour of the reactor facility. d) The Berkeley City Council began a study of problems associated with radioactive material within the city. e) Friends Of The Earth formally petitionedmore » the NRC to terminate the reactor's license. Campus personnel have expended many man-hours and many pounds of paper in responding to these happenings. Some of the details are of interest, and may be of use to other reactor facilities. (author)« less

Decommissioning the physics laboratory, building 777-10A, at the Savannah River Site (SRS)

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

Musall, John C.; Cope, Jeff L.

2008-01-15

SRS recently completed a four year mission to decommission {approx}250 excess facilities. As part of that effort, SRS decommissioned a 48,000 ft{sup 2} laboratory that housed four low-power test reactors, formerly used by SRS to determine reactor physics. This paper describes and reviews the decommissioning, with a focus on component segmentation and handling (i.e. hazardous material removal, demolition, and waste handling). The paper is intended to be a resource for engineers, planners, and project managers, who face similar decommissioning challenges. Building 777-10A, located at the south end of SRS's A/M-Area, was built in 1953 and had a gross area of {approx}48,000 ft{sup 2}. Building 777-10A had two main areas: a west wing, which housed four experimental reactors and associated equipment; and an east wing, which housed laboratories, and shops, offices. The reactors were located in two separate areas: one area housed the Process Development Pile (PDP) reactor and the Lattice Test Reactor (LTR), while the second area housed the Standard Pile (SP) and the Sub-critical Experiment (SE) reactors. The west wing had five levels: three below and three above grade (floor elevations of -37', -28', -15', 0', +13'/+16' and +27' (roof elevation of +62')), while the east wing had two levels: one below and one above grade (floor elevations of -15' and 0' (roof elevation of +16')). Below-grade exterior walls were constructed of reinforced concrete, {approx}1' thick. In general, above-grade exterior walls were steel frames covered by insulation and corrugated, asbestos-cement board. The two interior walls around the PDP/LTR were reinforced concrete {approx}5' thick and {approx}30' high, while the SP/SE reactors resided in a reinforced, concrete cell with 3.5'-6' thick walls/roof. All other interior walls were constructed of metal studs covered with either asbestos-cement or gypsum board. In general, the floors were constructed of reinforced concrete on cast-in-place concrete beams below-grade and concrete on metal beams above-grade. The roofs were flat concrete slabs on metal beams. Building 777-10A was an important SRS research and development location. The reactors helped determine safe operational limits and loading patterns for fuel used in the SRS production reactors, and supported various low power reactor physics studies. All four reactors were shut down and de-inventoried in the 1970's. The building was DD and R 2007, Chattanooga, Tennessee, September 16-19, 2007 169 subsequently used by various SRS organizations for office space, audio/visual studio, and computer network hub. SRS successfully decommissioned Building 777-10A over a thirty month period at a cost of {approx}more » $$14 M ({approx}$$290/ft{sup 2}). The decommissioning was a complex and difficult effort due to the building's radiological contamination, height, extensive basement, and thick concrete walls. Extensive planning and extensive hazard analysis (e.g. of structural loads/modifications leading to unplanned collapse) ensured the decommissioning was completed safely and without incident. The decommissioning met contract standards for residual contamination and physical/chemical hazards, and was the last in a series of decommissioning projects that prepared the lower A/M-Area for SRS's environmental restoration program.« less

Small low mass advanced PBR's for propulsion

NASA Astrophysics Data System (ADS)

Powell, J. R.; Todosow, M.; Ludewig, H.

1993-10-01

The advanced Particle Bed Reactor (PBR) to be described in this paper is characterized by relatively low power, and low cost, while still maintaining competition values for thrust/weight, specific impulse and operating times. In order to retain competitive values for the thrust/weight ratio while reducing the reactor size, it is necessary to change the basic reactor layout, by incorporating new concepts. The new reactor design concept is termed SIRIUS (Small Lightweight Reactor Integral Propulsion System). The following modifications are proposed for the reactor design to be discussed in this paper: Pre-heater (U-235 included in Moderator); Hy-C (Hydride/De-hydride for Reactor Control); Afterburner (U-235 impregnated into Hot Frit); and Hy-S (Hydride Spike Inside Hot Frit). Each of the modifications will be briefly discussed below, with benefits, technical issues, design approach, and risk levels addressed. The paper discusses conceptual assumptions, feasibility analysis, mass estimates, and information needs.

Physics and nuclear power

NASA Astrophysics Data System (ADS)

Buttery, N. E.

2008-03-01

Nuclear power owes its origin to physicists. Fission was demonstrated by physicists and chemists and the first nuclear reactor project was led by physicists. However as nuclear power was harnessed to produce electricity the role of the engineer became stronger. Modern nuclear power reactors bring together the skills of physicists, chemists, chemical engineers, electrical engineers, mechanical engineers and civil engineers. The paper illustrates this by considering the Sizewell B project and the role played by physicists in this. This covers not only the roles in design and analysis but in problem solving during the commissioning of first of a kind plant. Looking forward to the challenges to provide sustainable and environmentally acceptable energy sources for the future illustrates the need for a continuing synergy between physics and engineering. This will be discussed in the context of the challenges posed by Generation IV reactors.

Demand driven salt clean-up in a molten salt fast reactor - Defining a priority list.

PubMed

Merk, B; Litskevich, D; Gregg, R; Mount, A R

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified.

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

Bess, John D.; Sterbentz, James W.; Snoj, Luka

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Physics From the News -- Fukushima Daiichi: Radiation Doses and Dose Rates

NASA Astrophysics Data System (ADS)

Bartlett, A. A.

2011-09-01

The nuclear disaster that was triggered by the Japanese earthquake and the following tsunami of March 11, 2011, continues to be the subject of a great deal of news coverage. The tsunami caused severe damage to the nuclear power reactors at Fukushima Daiichi, and this led to the escape of unknown quantities of radioactive material from the damaged fuel rods in the reactors and from the associated storage facilities for the fuel rods that had been removed from the reactors.

Health Physics State of the RA-1 Reactor, Period Between 1-1-62 and 10-8- 62. Report No. 87; ESTADO RADIOSANIT ARIO DEL REACTOR R.A. 1, PERIODO COMPRENDIDO ENTRE EL 1-1-62 Y EL 10-8-62. Informe No. 87

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

Gaspar, R.; Moll, O.; Hermelo, C.

1963-01-01

The methods used to measure the irradiation levels and the radiation exposure of personnel of the RA-1 reactor are described. The criteria used to evaluate the risks from this exposure are reported. Typical graphs are shown of the radiation levels measured in the control room. (J.S.R.)

Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

NASA Technical Reports Server (NTRS)

Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

1991-01-01

The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

Synfuels from fusion: using the tandem mirror reactor and a thermochemical cycle to produce hydrogen

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

Werner, R.W.

1982-11-01

This study is concerned with the following area: (1) the tandem mirror reactor and its physics; (2) energy balance; (3) the lithium oxide canister blanket system; (4) high-temperature blanket; (5) energy transport system-reactor to process; (6) thermochemical hydrogen processes; (7) interfacing the GA cycle; (8) matching power and temperature demands; (9) preliminary cost estimates; (10) synfuels beyond hydrogen; and (11) thermodynamics of the H/sub 2/SO/sub 4/-H/sub 2/O system. (MOW)

Horizontal baffle for nuclear reactors

DOEpatents

Rylatt, John A.

1978-01-01

A horizontal baffle disposed in the annulus defined between the core barrel and the thermal liner of a nuclear reactor thereby physically separating the outlet region of the core from the annular area below the horizontal baffle. The horizontal baffle prevents hot coolant that has passed through the reactor core from thermally damaging apparatus located in the annulus below the horizontal baffle by utilizing the thermally induced bowing of the horizontal baffle to enhance sealing while accommodating lateral motion of the baffle base plate.

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

Steady State Advanced Tokamak (SSAT): The mission and the machine

NASA Astrophysics Data System (ADS)

Thomassen, K.; Goldston, R.; Nevins, B.; Neilson, H.; Shannon, T.; Montgomery, B.

1992-03-01

Extending the tokamak concept to the steady state regime and pursuing advances in tokamak physics are important and complementary steps for the magnetic fusion energy program. The required transition away from inductive current drive will provide exciting opportunities for advances in tokamak physics, as well as important impetus to drive advances in fusion technology. Recognizing this, the Fusion Policy Advisory Committee and the U.S. National Energy Strategy identified the development of steady state tokamak physics and technology, and improvements in the tokamak concept, as vital elements in the magnetic fusion energy development plan. Both called for the construction of a steady state tokamak facility to address these plan elements. Advances in physics that produce better confinement and higher pressure limits are required for a similar unit size reactor. Regimes with largely self-driven plasma current are required to permit a steady-state tokamak reactor with acceptable recirculating power. Reliable techniques of disruption control will be needed to achieve the availability goals of an economic reactor. Thus the central role of this new tokamak facility is to point the way to a more attractive demonstration reactor (DEMO) than the present data base would support. To meet the challenges, we propose a new 'Steady State Advanced Tokamak' (SSAT) facility that would develop and demonstrate optimized steady state tokamak operating mode. While other tokamaks in the world program employ superconducting toroidal field coils, SSAT would be the first major tokamak to operate with a fully superconducting coil set in the elongated, divertor geometry planned for ITER and DEMO.

Neutron time-of-flight spectroscopy measurement using a waveform digitizer

NASA Astrophysics Data System (ADS)

Liu, Long-Xiang; Wang, Hong-Wei; Ma, Yu-Gang; Cao, Xi-Guang; Cai, Xiang-Zhou; Chen, Jin-Gen; Zhang, Gui-Lin; Han, Jian-Long; Zhang, Guo-Qiang; Hu, Ji-Feng; Wang, Xiao-He

2016-05-01

The photoneutron source (PNS, phase 1), an electron linear accelerator (linac)-based pulsed neutron facility that uses the time-of-flight (TOF) technique, was constructed for the acquisition of nuclear data from the Thorium Molten Salt Reactor (TMSR) at the Shanghai Institute of Applied Physics (SINAP). The neutron detector signal used for TOF calculation, with information on the pulse arrival time, pulse shape, and pulse height, was recorded by using a waveform digitizer (WFD). By using the pulse height and pulse-shape discrimination (PSD) analysis to identify neutrons and γ-rays, the neutron TOF spectrum was obtained by employing a simple electronic design, and a new WFD-based DAQ system was developed and tested in this commissioning experiment. The DAQ system developed is characterized by a very high efficiency with respect to millisecond neutron TOF spectroscopy. Supported by Strategic Priority Research Program of the Chinese Academy of Science(TMSR) (XDA02010100), National Natural Science Foundation of China(NSFC)(11475245,No.11305239), Shanghai Key Laboratory of Particle Physics and Cosmology (11DZ2260700)

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

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

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

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.

Sensitivity to VSR failure: K pipe break accident

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

Meichle, R.H.

1969-09-12

Reactor effects of failure of a safety rod to scram can be considered in two major respects: The reduction in total safety system strength which will affect the amount of ``prompt drop`` and subsequent flux decay rate of the average neutron flux-level; and the change in local flux distribution due to the absence of the particular rod which fails to enter the reactor. The purpose of this memorandum is to describe the physical effects involved and to indicate the approximate magnitude of both reactor-wide and localized changes in event of failure of a VSR simultaneous with a K Reactor risermore » accident.« less

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.

Methods of conducting simultaneous exothermic and endothermic reactions

DOEpatents

Tonkovich, Anna Lee [Marysville, OH; Roberts, Gary L [West Richland, WA; Perry, Steven T [Galloway, OH; Fitzgerald, Sean P [Columbus, OH

2005-11-29

Integrated Combustion Reactors (ICRs) and methods of making ICRs are described in which combustion chambers (or channels) are in direct thermal contact to reaction chambers for an endothermic reaction. Superior results were achieved for combustion chambers which contained a gap for free flow through the chamber. Particular reactor designs are also described. Processes of conducting reactions in integrated combustion reactors are described and results presented. Some of these processes are characterized by unexpected and superior results.

Acquisition of Raman Spectrometer and High Temperature and Pressure Reactor for Synthesis and Characterization of Carbon Based Hybrid Nanoparticles from Waste Wood

DTIC Science & Technology

2015-04-27

from waste biomass using these two high temperature reactors. We have extensively used a Raman spectrometer to analyse as synthesized carbon materials...corporation). These tools were fully installed and operational. We have also synthesized carbon materials from waste biomass using these two high...materials from waste biomass using these two high temperature reactors. We have extensively used a Raman spectrometer to analyse as synthesized carbon

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

NASA Astrophysics Data System (ADS)

Marshalkin, V. E.; Povyshev, V. M.

2015-12-01

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.

Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Mangiarotti, F.; Minervini, J.

2014-10-01

The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

Uranium to Electricity: The Chemistry of the Nuclear Fuel Cycle

ERIC Educational Resources Information Center

Settle, Frank A.

2009-01-01

The nuclear fuel cycle consists of a series of industrial processes that produce fuel for the production of electricity in nuclear reactors, use the fuel to generate electricity, and subsequently manage the spent reactor fuel. While the physics and engineering of controlled fission are central to the generation of nuclear power, chemistry…

RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

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

David Andrs; Ray Berry; Derek Gaston

The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7more » is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to evolve with time. RELAP-7 is a MOOSE-based application. MOOSE (Multiphysics Object-Oriented Simulation Environment) is a framework for solving computational engineering problems in a well-planned, managed, and coordinated way. By leveraging millions of lines of open source software packages, such as PETSC (a nonlinear solver developed at Argonne National Laboratory) and LibMesh (a Finite Element Analysis package developed at University of Texas), MOOSE significantly reduces the expense and time required to develop new applications. Numerical integration methods and mesh management for parallel computation are provided by MOOSE. Therefore RELAP-7 code developers only need to focus on physics and user experiences. By using the MOOSE development environment, RELAP-7 code is developed by following the same modern software design paradigms used for other MOOSE development efforts. There are currently over 20 different MOOSE based applications ranging from 3-D transient neutron transport, detailed 3-D transient fuel performance analysis, to long-term material aging. Multi-physics and multiple dimensional analyses capabilities can be obtained by coupling RELAP-7 and other MOOSE based applications and by leveraging with capabilities developed by other DOE programs. This allows restricting the focus of RELAP-7 to systems analysis-type simulations and gives priority to retain and significantly extend RELAP5's capabilities.« less

Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

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

Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.

Here, we present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after theymore » pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.« less

Plasma flow reactor for steady state monitoring of physical and chemical processes at high temperatures

DOE PAGES

Koroglu, Batikan; Mehl, Marco; Armstrong, Michael R.; ...

2017-09-11

Here, we present the development of a steady state plasma flow reactor to investigate gas phase physical and chemical processes that occur at high temperature (1000 < T < 5000 K) and atmospheric pressure. The reactor consists of a glass tube that is attached to an inductively coupled argon plasma generator via an adaptor (ring flow injector). We have modeled the system using computational fluid dynamics simulations that are bounded by measured temperatures. In situ line-of-sight optical emission and absorption spectroscopy have been used to determine the structures and concentrations of molecules formed during rapid cooling of reactants after theymore » pass through the plasma. Emission spectroscopy also enables us to determine the temperatures at which these dynamic processes occur. A sample collection probe inserted from the open end of the reactor is used to collect condensed materials and analyze them ex situ using electron microscopy. The preliminary results of two separate investigations involving the condensation of metal oxides and chemical kinetics of high-temperature gas reactions are discussed.« less

CHLORINE ABSORPTION IN S(IV) SOLUTIONS

EPA Science Inventory

The report gives results of measurements of the rate of Chlorine (Cl2) absorption into aqueous sulfite/bisulfite -- S(IV) -- solutions at ambient temperature using a highly characterized stirred-cell reactor. The reactor media were 0 to 10 mM S(IV) with pHs of 3.5-8.5. Experiment...

Nonlinear Ultrasonic Measurements in Nuclear Reactor Environments

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this thesis, two ultrasonic characterization techniques will be explored. The first, finite amplitude wave propagation has been demonstrated to be sensitive to microstructural material property changes. It is a strong candidate to determine fuel evolution; however, it has not been demonstrated for in-situ reactor applications. In this thesis, finite amplitude wave propagation will be used to measure the microstructural evolution in Al-6061. This is the first demonstration of finite amplitude wave propagation at temperatures in excess of 200 °C and during an irradiation test. Second, a method based on contact nonlinear acoustic theory will be developed to identify compressed cracks. Compressed cracks are typically transparent to ultrasonic wave propagation; however, by measuring harmonic content developed during finite amplitude wave propagation, it is shown that even compressed cracks can be characterized. Lastly, piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65x10 20 n/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 x1020 n/cm2, Zinc Oxide is capable of transduction up to 6.27 x1020 n/cm 2, and Aluminum Nitride is capable of transduction up to 8.65x x10 20 n/cm2.

Study of gain homogeneity and radiation effects of Low Gain Avalanche Pad Detectors

NASA Astrophysics Data System (ADS)

Gallrapp, C.; Fernández García, M.; Hidalgo, S.; Mateu, I.; Moll, M.; Otero Ugobono, S.; Pellegrini, G.

2017-12-01

Silicon detectors with intrinsic charge amplification implementing a n++-p+-p structure are considered as a sensor technology for future tracking and timing applications in high energy physics experiments. The performance of the intrinsic gain in Low Gain Avalanche Detectors (LGAD) after irradiation is crucial for the characterization of radiation hardness and timing properties in this technology. LGAD devices irradiated with reactor neutrons or 800 MeV protons reaching fluences of 2.3 × 1016 neq/cm2 were characterized using Transient Current Technique (TCT) measurements with red and infra-red laser pulses. Leakage current variations observed in different production lots and within wafers were investigated using Thermally Stimulated Current (TSC). Results showed that the intrinsic charge amplification is reduced with increasing fluence up to 1015 neq/cm2 which is related to an effective acceptor removal. Further relevant issues were charge collection homogeneity across the detector surface and leakage current performance before and after irradiation.

Sustainable Thorium Nuclear Fuel Cycles: A Comparison of Intermediate and Fast Neutron Spectrum Systems

DOE PAGES

Brown, Nicholas R.; Powers, Jeffrey J.; Feng, B.; ...

2015-05-21

This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight latticemore » heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this selfsustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.« less

Characterization and anaerobic treatment of the sanitary landfill leachate in Istanbul.

PubMed

Inanc, B; Calli, B; Saatci, A

2000-01-01

In this study, characterization and anaerobic treatability of leachate from Komurcuoda Sanitary Landfill located on the Asian part of Istanbul were investigated. Time based fluctuations in characteristics of leachate were monitored for an 8 month period. Samples were taken from a 200 m3 holding tank located at the lowest elevation of the landfill. COD concentrations have ranged between 18,800 and 47,800 mg/l while BOD5 between 6820 and 38,500 mg/L. COD and BOD5 values were higher in summer and lower in winter due to dilution by precipitation. On the other hand, it was quite interesting that such a dilution effect was not observed for ammonia. The highest ammonia concentration, 2690 mg/L was in November 1998. BOD5/COD ratio was larger than 0.7 for most samples indicating high biodegradability, and acidic phase of decomposition in the landfill. For anaerobic treatability, three different reactors, namely an upflow anaerobic sludge bed reactor, an anaerobic upflow filter and a hybrid bed reactor, were used. The anaerobic reactors were operated for more than 230 days and were continuing operation when this paper was prepared. Organic loading was increased gradually from 1.3 kg COD/m3.day to 8.2 kg COD/m3.day while hydraulic retention time was reduced from 2.4 days to 2.0 days. All the reactors showed similar performances against organic loadings with efficiencies between 80% and 90%. However the reactors have experienced high ammonia concentrations several times throughout the experimental period, and showed different inhibition levels. Anaerobic filter was the least affected reactor while UASB was the most. Hybrid bed reactor has exhibited a similar performance to anaerobic filter although not to the same degree.

Corrigendum to “Atom probe tomography characterization of neutron irradiated surveillance samples from the R.E. Ginna reactor pressure vessel”

DOE PAGES

Edmondson, Philip D.; Miller, Michael K.; Powers, K. A.; ...

2017-03-24

In our recent paper entitled “Atom probe tomography characterization of neutron irradiated surveillance samples from the R. E. Ginna reactor pressure vessel”, we make reference to a table within the article as providing the average compositions of the precipitates, when in fact the bulk compositions were given. In this correction, we present the average precipitate compositions for the data presented in Ref. [1]. These correct compositions are provided for information and do not alter the conclusions of the original manuscript.

Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

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

Jensen, C.; Wachs, D.; Carmack, J.

The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, andmore » salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.« less

Vibro-acoustic Imaging at the Breazeale Reactor

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

Smith, James Arthur; Jewell, James Keith; Lee, James Edwin

2016-09-01

The INL is developing Vibro-acoustic imaging technology to characterize microstructure in fuels and materials in spent fuel pools and within reactor vessels. A vibro-acoustic development laboratory has been established at the INL. The progress in developing the vibro-acoustic technology at the INL is the focus of this report. A successful technology demonstration was performed in a working TRIGA research reactor. Vibro-acoustic imaging was performed in the reactor pool of the Breazeale reactor in late September of 2015. A confocal transducer driven at a nominal 3 MHz was used to collect the 60 kHz differential beat frequency induced in a spentmore » TRIGA fuel rod and empty gamma tube located in the main reactor water pool. Data was collected and analyzed with the INLDAS data acquisition software using a short time Fourier transform.« less

Portable vibro-acoustic testing system for in situ microstructure characterization and metrology

NASA Astrophysics Data System (ADS)

Smith, James A.; Nichol, Corrie I.; Zuck, Larry D.; Fatemi, Mostafa

2018-04-01

There is a need in research reactors like the one at INL to inspect irradiated materials and structures. The goal of this work is to develop a portable scanning infrastructure for a material characterization technique called vibro-acoustography (VA) that has been developed by the Idaho National laboratory for nuclear applications to characterize fuel, cladding materials, and structures. The proposed VA technology is based on ultrasound and acoustic waves; however, it provides information beyond what is available from the traditional ultrasound techniques and can expand the knowledge on nuclear material characterization and microstructure evolution. This paper will report on the development of a portable scanning system that will be set up to characterize materials and components in open water reactors and canals in situ. We will show some initial laboratory results of images generated by vibro-acoustics of surrogate fuel plates and graphite structures and discuss the design of the portable system.

Development and Characterization of 6Li-doped Liquid Scintillator Detectors for PROSPECT

NASA Astrophysics Data System (ADS)

Gaison, Jeremy; Prospect Collaboration

2016-09-01

PROSPECT, the Precision Reactor Oscillation and Spectrum experiment, is a phased reactor antineutrino experiment designed to search for eV-scale sterile neutrinos via short-baseline neutrino oscillations and to make a precision measurement of the 235U reactor antineutrino spectrum. A multi-ton, optically segmented detector will be deployed at Oak Ridge National Laboratory's (ORNL) High Flux Isotope Reactor (HFIR) to measure the reactor spectrum at baselines ranging from 7-12m. A two-segment detector prototype with 50 liters of active liquid scintillator target has been built to verify the detector design and to benchmark its performance. In this presentation, we will summarize the performance of this detector prototype and describe the optical and energy calibration of the segmented PROSPECT detectors.

Reactor design rules for GaN epitaxial layer growths on sapphire in metal-organic chemical vapour deposition

NASA Astrophysics Data System (ADS)

Kim, Keunjoo; Noh, Sam Kyu

2000-08-01

The thermal process of the growth of GaN-based semiconductors was analysed for two home-made horizontal reactors. The reactors were designed to make the ammonia gas flow in the opposite direction to the main gas flow. For two horizontal reactors different in dimension, the low Reynolds numbers of Re = 2.94 and 4.15 were chosen for stable laminar flow and the Rayleigh numbers governing the heat convection were optimized to the values of Ra = 6.0 and 76.2, respectively. The qualities of GaN and InGaN films were characterized by Hall effect measurement, x-ray diffraction and photoluminescence and compared with respect to the reactor dependency.

Fission neutron source in Rome

NASA Astrophysics Data System (ADS)

Coppola, Mario; Di Majo, V.; Ingrao, G.; Rebessi, S.; Testa, A.

1997-02-01

A fission neutron source is operating in Rome at the ENEA Casaccia Research Center since 1971, consisting of a low power fast reactor named RSV-Tapiro. it is employed for a variety of experiments, including dosimetry, material testing, radiation protection and biology. In particular, application to experimental radiobiology includes studies of the biological action of neutrons in the whole-body irradiated animal, or in specialized systems in vivo or in vitro. For his purpose a vertical irradiation facility was originally constructed. Recently, a new horizontal irradiation facility has been designed to allow the exposure of larger samples or larger sample batches at one time. Dosimetry at the sample irradiation positions is routinely carried out by the conventional method of using two ion chambers. This physical dosimetry has recently been compared with the results of biological dosimetry based on the detection of chromosomal aberrations in peripheral blood human lymphocytes irradiated in vitro. A characterization of the radiation quality in the two configurations has been carried out by tissue equivalent proportional counter microdosimetry measurements. Information about the main characteristics of the reactor and the two irradiation facilities is provided and relevant results of the various measurements are summarized. Radiobiological results obtained using this neutron source are also briefly outlined.

Responses of biofilm characteristics to variations in temperature and NH4(+)-N loading in a moving-bed biofilm reactor treating micro-polluted raw water.

PubMed

Zhang, Shuangfu; Wang, Yayi; He, Weitao; Wu, Min; Xing, Meiyan; Yang, Jian; Gao, Naiyun; Yin, Daqiang

2013-03-01

A pilot-scale moving-bed biofilm reactor (MBBR) for biological treatment of micro-polluted raw water was operated over 400days to investigate the responses of biofilm characteristics and nitrification performance to variations in temperature and NH4(+)-N loading. The mean removal efficiency of NH4(+)-N in the MBBR reached 71.4±26.9%, and batch experiments were performed to study nitrification kinetics for better process understanding. Seven physical-chemical parameters, including volatile solids (VS), polysaccharides (PS) and phospholipids (PL) increased firstly, and then rapidly decreased with increasing temperature and NH4(+)-N loading, and properly characterized the attached biomass during biofilm development and detachment in the MBBR. The biofilm compositions were described by six ratios, e.g., PS/VS and PL/VS ratios showed different variation trends, indicating different responses of PS and PL to the changes in temperature and NH4(+)-N loading. Furthermore, fluorescent in situ hybridization (FISH) analysis revealed that increased NH4(+)-N loadings caused an enrichment of the nitrifying biofilm. Copyright © 2013 Elsevier Ltd. All rights reserved.

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA from use at their institutions. These areas are discussed in this report.« less

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

NONE

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.« less

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

Mulder, R.U.; Benneche, P.E.; Hosticka, B.

The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would notmore » be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.« less

A PRECISION MEASUREMENT OF THE NEUTRINO MIXING ANGLE THETA (SUB 13) USING REACTOR ANTINEUTRINOS AT DAYA BAY.

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

KETTELL, S.; ET AL.

2006-10-16

This document describes the design of the Daya Bay reactor neutrino experiment. Recent discoveries in neutrino physics have shown that the Standard Model of particle physics is incomplete. The observation of neutrino oscillations has unequivocally demonstrated that the masses of neutrinos are nonzero. The smallness of the neutrino masses (<2 eV) and the two surprisingly large mixing angles measured have thus far provided important clues and constraints to extensions of the Standard Model. The third mixing angle, {delta}{sub 13}, is small and has not yet been determined; the current experimental bound is sin{sup 2} 2{theta}{sub 13} < 0.17 at 90%more » confidence level (from Chooz) for {Delta}m{sub 31}{sup 2} = 2.5 x 10{sup -3} eV{sup 2}. It is important to measure this angle to provide further insight on how to extend the Standard Model. A precision measurement of sin{sup 2} 2{theta}{sub 13} using nuclear reactors has been recommended by the 2004 APS Multi-divisional Study on the Future of Neutrino Physics as well as a recent Neutrino Scientific Assessment Group (NUSAG) report. We propose to perform a precision measurement of this mixing angle by searching for the disappearance of electron antineutrinos from the nuclear reactor complex in Daya Bay, China. A reactor-based determination of sin{sup 2} 2{theta}{sub 13} will be vital in resolving the neutrino-mass hierarchy and future measurements of CP violation in the lepton sector because this technique cleanly separates {theta}{sub 13} from CP violation and effects of neutrino propagation in the earth. A reactor-based determination of sin{sup 2} 2{theta}{sub 13} will provide important, complementary information to that from long-baseline, accelerator-based experiments. The goal of the Daya Bay experiment is to reach a sensitivity of 0.01 or better in sin{sup 2} 2{theta}{sub 13} at 90% confidence level.« less

LWR pressure vessel surveillance dosimetry improvement program: LWR power reactor surveillance physics-dosimetry data base compendium

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

McElroy, W.N.

1985-08-01

This NRC physics-dosimetry compendium is a collation of information and data developed from available research and commercial light water reactor vessel surveillance program (RVSP) documents and related surveillance capsule reports. The data represents the results of the HEDL least-squares FERRET-SAND II Code re-evaluation of exposure units and values for 47 PWR and BWR surveillance capsules for W, B and W, CE, and GE power plants. Using a consistent set of auxiliary data and dosimetry-adjusted reactor physics results, the revised fluence values for E > 1 MeV averaged 25% higher than the originally reported values. The range of fluence values (new/old)more » was from a low of 0.80 to a high of 2.38. These HEDL-derived FERRET-SAND II exposure parameter values are being used for NRC-supported HEDL and other PWR and BWR trend curve data development and testing studies. These studies are providing results to support Revision 2 of Regulatory Guide 1.99. As stated by Randall (Ra84), the Guide is being updated to reflect recent studies of the physical basis for neutron radiation damage and efforts to correlate damage to chemical composition and fluence.« less

The search for sterile neutrinos at reactors and underground laboratories

NASA Astrophysics Data System (ADS)

Langford, Thomas

2017-01-01

From the initial discovery of neutrinos to the observation of neutrino oscillations, unexpected results have lead to deeper understanding of physics. However, as experiments and theoretical predictions have improved, new anomalies have surfaced that could point to beyond the Standard Model physics. Leading hypotheses invoke a new form of matter, sterile neutrinos, as a possible resolution of these outstanding questions. New experimental efforts are underway to probe short-baseline neutrino oscillations with reactors and radioactive sources. This talk will highlight developments in current and next generation experiments and present possible outcomes for the next few years.

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.

Nuclear and Physical Properties of Dielectrics under Neutron Irradiation in Fast (BN-600) and Fusion (DEMO-S) Reactors

NASA Astrophysics Data System (ADS)

Blokhin, D. A.; Chernov, V. M.; Blokhin, A. I.

2017-12-01

Nuclear and physical properties (activation and transmutation of elements) of BN and Al2O3 dielectric materials subjected to neutron irradiation for up to 5 years in Russian fast (BN-600) and fusion (DEMO-S) reactors were calculated using the ACDAM-2.0 software complex for different post-irradiation cooling times (up to 10 years). Analytical relations were derived for the calculated quantities. The results may be used in the analysis of properties of irradiated dielectric materials and may help establish the rules for safe handling of these materials.

Alternate fusion fuels workshop

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

Not Available

1981-06-01

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

2009-08-01

This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

Code of Federal Regulations, 2013 CFR

2013-01-01

... shall: (i) Design, construct, install and maintain physical barriers as necessary to control access into.... (10) Vehicle control measures. Consistent with the physical protection program design requirements of... maintain vehicle control measures, as necessary, to protect against the design basis threat of radiological...

10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

Code of Federal Regulations, 2012 CFR

2012-01-01

... shall: (i) Design, construct, install and maintain physical barriers as necessary to control access into.... (10) Vehicle control measures. Consistent with the physical protection program design requirements of... maintain vehicle control measures, as necessary, to protect against the design basis threat of radiological...

10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

Code of Federal Regulations, 2014 CFR

2014-01-01

... shall: (i) Design, construct, install and maintain physical barriers as necessary to control access into.... (10) Vehicle control measures. Consistent with the physical protection program design requirements of... maintain vehicle control measures, as necessary, to protect against the design basis threat of radiological...

Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

NASA Astrophysics Data System (ADS)

Zhou, Hao-Jun; Yin, Yan-Peng; Fan, Xiao-Qiang; Li, Zheng-Hong; Pu, Yi-Kang

2016-06-01

A perturbation method is proposed to obtain the effective delayed neutron fraction β eff of a cylindrical highly enriched uranium reactor. Based on reactivity measurements with and without a sample at a specified position using the positive period technique, the reactor reactivity perturbation Δρ of the sample in β eff units is measured. Simulations of the perturbation experiments are performed using the MCNP program. The PERT card is used to provide the difference dk of effective neutron multiplication factors with and without the sample inside the reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation β eff = dk/Δρ is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated. The average β eff value of the reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for β eff can be used to evaluate the reliabilities of the delayed neutron parameters. This work shows that the delayed neutron data of 235U and 238U from G.R. Keepin’s publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2. Supported by Foundation of Key Laboratory of Neutron Physics, China Academy of Engineering Physics (2012AA01, 2014AA01), National Natural Science Foundation (11375158, 91326104)

Research and proposal on selective catalytic reduction reactor optimization for industrial boiler.

PubMed

Yang, Yiming; Li, Jian; He, Hong

2017-08-24

The advanced computational fluid dynamics (CFD) software STAR-CCM+ was used to simulate a denitrification (De-NOx) project for a boiler in this paper, and the simulation result was verified based on a physical model. Two selective catalytic reduction (SCR) reactors were developed: reactor 1 was optimized and reactor 2 was developed based on reactor 1. Various indicators, including gas flow field, ammonia concentration distribution, temperature distribution, gas incident angle, and system pressure drop were analyzed. The analysis indicated that reactor 2 was of outstanding performance and could simplify developing greatly. Ammonia injection grid (AIG), the core component of the reactor, was studied; three AIGs were developed and their performances were compared and analyzed. The result indicated that AIG 3 was of the best performance. The technical indicators were proposed for SCR reactor based on the study. Flow filed distribution, gas incident angle, and temperature distribution are subjected to SCR reactor shape to a great extent, and reactor 2 proposed in this paper was of outstanding performance; ammonia concentration distribution is subjected to ammonia injection grid (AIG) shape, and AIG 3 could meet the technical indicator of ammonia concentration without mounting ammonia mixer. The developments above on the reactor and the AIG are both of great application value and social efficiency.

Bacterial Colonization of Pellet Softening Reactors Used during Drinking Water Treatment▿

PubMed Central

Hammes, Frederik; Boon, Nico; Vital, Marius; Ross, Petra; Magic-Knezev, Aleksandra; Dignum, Marco

2011-01-01

Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase the pH and facilitate the process of precipitation on an added seeding material. Here we describe for the first time the opportunistic bacterial colonization of the calcite pellets in a full-scale pellet softening reactor and the functional contribution of these colonizing bacteria to the overall drinking water treatment process. ATP analysis, advanced microscopy, and community fingerprinting with denaturing gradient gel electrophoretic (DGGE) analysis were used to characterize the biomass on the pellets, while assimilable organic carbon (AOC), dissolved organic carbon, and flow cytometric analysis were used to characterize the impact of the biological processes on drinking water quality. The data revealed pellet colonization at concentrations in excess of 500 ng of ATP/g of pellet and reactor biomass concentrations as high as 220 mg of ATP/m3 of reactor, comprising a wide variety of different microorganisms. These organisms removed as much as 60% of AOC from the water during treatment, thus contributing toward the biological stabilization of the drinking water. Notably, only a small fraction (about 60,000 cells/ml) of the bacteria in the reactors was released into the effluent under normal conditions, while the majority of the bacteria colonizing the pellets were captured in the calcite structures of the pellets and were removed as a reusable product. PMID:21148700

Bacterial colonization of pellet softening reactors used during drinking water treatment.

PubMed

Hammes, Frederik; Boon, Nico; Vital, Marius; Ross, Petra; Magic-Knezev, Aleksandra; Dignum, Marco

2011-02-01

Pellet softening reactors are used in centralized and decentralized drinking water treatment plants for the removal of calcium (hardness) through chemically induced precipitation of calcite. This is accomplished in fluidized pellet reactors, where a strong base is added to the influent to increase the pH and facilitate the process of precipitation on an added seeding material. Here we describe for the first time the opportunistic bacterial colonization of the calcite pellets in a full-scale pellet softening reactor and the functional contribution of these colonizing bacteria to the overall drinking water treatment process. ATP analysis, advanced microscopy, and community fingerprinting with denaturing gradient gel electrophoretic (DGGE) analysis were used to characterize the biomass on the pellets, while assimilable organic carbon (AOC), dissolved organic carbon, and flow cytometric analysis were used to characterize the impact of the biological processes on drinking water quality. The data revealed pellet colonization at concentrations in excess of 500 ng of ATP/g of pellet and reactor biomass concentrations as high as 220 mg of ATP/m(3) of reactor, comprising a wide variety of different microorganisms. These organisms removed as much as 60% of AOC from the water during treatment, thus contributing toward the biological stabilization of the drinking water. Notably, only a small fraction (about 60,000 cells/ml) of the bacteria in the reactors was released into the effluent under normal conditions, while the majority of the bacteria colonizing the pellets were captured in the calcite structures of the pellets and were removed as a reusable product.

An optically accessible pyrolysis microreactor

NASA Astrophysics Data System (ADS)

Baraban, J. H.; David, D. E.; Ellison, G. Barney; Daily, J. W.

2016-01-01

We report an optically accessible pyrolysis micro-reactor suitable for in situ laser spectroscopic measurements. A radiative heating design allows for completely unobstructed views of the micro-reactor along two axes. The maximum temperature demonstrated here is only 1300 K (as opposed to 1700 K for the usual SiC micro-reactor) because of the melting point of fused silica, but alternative transparent materials will allow for higher temperatures. Laser induced fluorescence measurements on nitric oxide are presented as a proof of principle for spectroscopic characterization of pyrolysis conditions.

An optically accessible pyrolysis microreactor

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

Baraban, J. H.; Ellison, G. Barney; David, D. E.

2016-01-15

We report an optically accessible pyrolysis micro-reactor suitable for in situ laser spectroscopic measurements. A radiative heating design allows for completely unobstructed views of the micro-reactor along two axes. The maximum temperature demonstrated here is only 1300 K (as opposed to 1700 K for the usual SiC micro-reactor) because of the melting point of fused silica, but alternative transparent materials will allow for higher temperatures. Laser induced fluorescence measurements on nitric oxide are presented as a proof of principle for spectroscopic characterization of pyrolysis conditions.

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

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

Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

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

J. Blair Briggs; John D. Bess; Jim Gulliford

2011-09-01

Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical ormore » subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the ICSBEP and the IRPhEP will be discussed in the full paper, selected benchmarks that have been added to the ICSBEP Handbook will be highlighted, and a preview of the new benchmarks that will appear in the September 2011 edition of the Handbook will be provided. Accomplishments of the IRPhEP will also be highlighted and the future of both projects will be discussed. REFERENCES (1) International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03/I-IX, Organisation for Economic Co-operation and Development-Nuclear Energy Agency (OECD-NEA), September 2010 Edition, ISBN 978-92-64-99140-8. (2) International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC(2006)1, Organisation for Economic Co-operation and Development-Nuclear Energy Agency (OECD-NEA), March 2011 Edition, ISBN 978-92-64-99141-5.« less

Issues relating to spent nuclear fuel storage on the Oak Ridge Reservation

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

Klein, J.A.; Turner, D.W.

1994-12-31

Currently, about 2,800 metric tons of spent nuclear fuel (SNF) is stored in the US, 1,000 kg of SNF (or about 0.03% of the nation`s total) are stored at the US Department of Energy (DOE) complex in Oak Ridge, Tennessee. However small the total quantity of material stored at Oak Ridge, some of the material is quite singular in character and, thus, poses unique management concerns. The various types of SNF stored at Oak Ridge will be discussed including: (1) High-Flux Isotope Reactor (HFIR) and future Advanced Neutron Source (ANS) fuels; (2) Material Testing Reactor (MTR) fuels, including Bulk Shieldingmore » Reactor (BSR) and Oak Ridge Research Reactor (ORR) fuels; (3) Molten Salt Reactor Experiment (MSRE) fuel; (4) Homogeneous Reactor Experiment (HRE) fuel; (5) Miscellaneous SNF stored in Oak Ridge National Laboratory`s (ORNL`s) Solid Waste Storage Areas (SWSAs); (6) SNF stored in the Y-12 Plant 9720-5 Warehouse including Health. Physics Reactor (HPRR), Space Nuclear Auxiliary Power (SNAP-) 10A, and DOE Demonstration Reactor fuels.« less

Characteristics of aerobic granules grown on glucose and acetate in sequential aerobic sludge blanket reactors.

PubMed

Tay, J H; Liu, Q S; Liu, Y

2002-08-01

Aerobic granules were cultivated in two column-type sequential aerobic sludge blanket reactors fed with glucose and acetate, respectively. The characteristics of aerobic granules were investigated. Results indicated that the glucose- and acetate-fed granules have comparable characteristics in terms of settling velocity, size, shape, biomass density, hydrophobicity, physical strength, microbial activity and storage stability. Substrate component does not seem to be a key factor on the formation of aerobic granules. However, microbial diversity of the granules is closely associated with the carbon sources supplied to the reactors. Compared with the conventional activated sludge flocs, aerobic granules exhibit excellent physical characteristics that would be essential for industrial application. This research provides a complete set of characteristics data of aerobic granules grown on glucose and acetate, which would be useful for further development of aerobic granules-based compact bioreactor for handling high strength organic wastewater.

American Nuclear Society 1994 student conference eastern region

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

NONE

This report contains abstracts from the 1994 American Nuclear Society Student Conference. The areas covered by these abstracts are: fusion and plasma physics; nuclear chemistry; radiation detection; reactor physics; thermal hydraulics; and corrosion science and waste issues.

Multivariate analysis of gamma spectra to characterize used nuclear fuel

DOE PAGES

Coble, Jamie; Orton, Christopher; Schwantes, Jon

2017-01-17

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Multivariate analysis of gamma spectra to characterize used nuclear fuel

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

Coble, Jamie; Orton, Christopher; Schwantes, Jon

The Multi-Isotope Process (MIP) Monitor provides an efficient means to monitor the process conditions in used nuclear fuel reprocessing facilities to support process verification and validation. The MIP Monitor applies multivariate analysis to gamma spectroscopy of key stages in the reprocessing stream in order to detect small changes in the gamma spectrum, which may indicate changes in process conditions. This research extends the MIP Monitor by characterizing a used fuel sample after initial dissolution according to the type of reactor of origin (pressurized or boiling water reactor; PWR and BWR, respectively), initial enrichment, burn up, and cooling time. Simulated gammamore » spectra were used in this paper to develop and test three fuel characterization algorithms. The classification and estimation models employed are based on the partial least squares regression (PLS) algorithm. A PLS discriminate analysis model was developed which perfectly classified reactor type for the three PWR and three BWR reactor designs studied. Locally weighted PLS models were fitted on-the-fly to estimate the remaining fuel characteristics. For the simulated gamma spectra considered, burn up was predicted with 0.1% root mean squared percent error (RMSPE) and both cooling time and initial enrichment with approximately 2% RMSPE. Finally, this approach to automated fuel characterization can be used to independently verify operator declarations of used fuel characteristics and to inform the MIP Monitor anomaly detection routines at later stages of the fuel reprocessing stream to improve sensitivity to changes in operational parameters that may indicate issues with operational control or malicious activities.« less

Characterization of elemental release during microbe-basalt interactions

NASA Astrophysics Data System (ADS)

Wu, L.; Jacobson, A. D.; Hausner, M.

2006-12-01

This study used batch reactors to characterize the rates, mechanisms, and stoichiometry of elemental release during the interaction of Burkholderia fungorum, a common soil microbe, with Columbia River Flood Basalt at 28°C for 36 d. We especially focused on the release of Ca, Mg, P, Si, and Sr under a variety of biotic and abiotic conditions with the ultimate aim of evaluating how actively metabolizing bacteria might influence basalt weathering on the continents. Four days after inoculating P-limited reactors (those lacking P in the growth medium), pH decreased from ~7 to 4, and glucose was depleted. Theoretical calculations suggest that the lowered pH resulted from the release of organic acids and/or CO2. Purely abiotic control reactors as well as control reactors containing nonviable cells showed constant glucose concentrations and near-neutral pH. Over the entire 36 day period, the P-limited reactors yielded Ca, Mg, Si, and Sr release rates several times higher than those observed in the P-bearing biotic reactors and the abiotic controls. Release rates directly correlate with pH, indicating that proton-promoted dissolution was the dominant reaction mechanism. Ligand- promoted dissolution was probably less important because the P-limited and P-bearing reactors experienced nearly identical rates of microbial growth, but the P-bearing reactors displayed overall lower dissolution rates at near-neutral pH, where presumably, the effect of ligand-promoted dissolution would be most evident. Chemical analyses of bacteria collected at the end of the experiments, combined with mass-balances between the biological and fluid phases, demonstrate that the low P concentration in the biotic reactors was an artifact of P uptake during microbial growth. These findings suggest that when bacteria utilize basalt as a nutrient source, they can potentially elevate the rate of long-term atmospheric CO2 consumption by Ca-Mg silicate weathering by a factor of 5 over the corresponding inorganic rate.

10 CFR 110.42 - Export licensing criteria.

Code of Federal Regulations, 2012 CFR

2012-01-01

... research on or development of any nuclear explosive device. (3) Adequate physical security measures will be... to exports of high-enriched uranium to be used as a fuel or target in a nuclear research or test... can be used in the reactor. (iii) A fuel or target “can be used” in a nuclear research or test reactor...

NUCLEAR REACTORS

DOEpatents

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

1961-12-01

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

Energy from the Atom. A Basic Teaching Unit on Energy. Revised.

ERIC Educational Resources Information Center

McDermott, Hugh, Ed.; Scharmann, Larry, Ed.

Recommended for grades 9-12 social studies and/or physical science classes, this 4-8 day unit focuses on four topics: (1) the background and history of atomic development; (2) two common types of nuclear reactors (boiling water and pressurized water reactors); (3) disposal of radioactive waste; and (4) the future of nuclear energy. Each topic…

Trickle-bed root culture bioreactor design and scale-up: growth, fluid-dynamics, and oxygen mass transfer.

PubMed

Ramakrishnan, Divakar; Curtis, Wayne R

2004-10-20

Trickle-bed root culture reactors are shown to achieve tissue concentrations as high as 36 g DW/L (752 g FW/L) at a scale of 14 L. Root growth rate in a 1.6-L reactor configuration with improved operational conditions is shown to be indistinguishable from the laboratory-scale benchmark, the shaker flask (mu=0.33 day(-1)). These results demonstrate that trickle-bed reactor systems can sustain tissue concentrations, growth rates and volumetric biomass productivities substantially higher than other reported bioreactor configurations. Mass transfer and fluid dynamics are characterized in trickle-bed root reactors to identify appropriate operating conditions and scale-up criteria. Root tissue respiration goes through a minimum with increasing liquid flow, which is qualitatively consistent with traditional trickle-bed performance. However, liquid hold-up is much higher than traditional trickle-beds and alternative correlations based on liquid hold-up per unit tissue mass are required to account for large changes in biomass volume fraction. Bioreactor characterization is sufficient to carry out preliminary design calculations that indicate scale-up feasibility to at least 10,000 liters.

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

Real-time monitoring of enzyme activity in a mesoporous silicon double layer

PubMed Central

Orosco, Manuel M.; Pacholski, Claudia; Sailor, Michael J.

2009-01-01

A double layer mesoporous silicon with different pore sizes functions as a nano-reactor that can isolate, filter and quantify the kinetics of enzyme reactions in real-time by optical reflectivity. This tiny reactor may be used to rapidly characterize a variety of isolated enzymes in a label-free manner. Activity of certain protease enzymes is often an indicator of disease states such as cancer1,2, stroke2, and neurodegeneracy3, and thus, there is a need for rapid assays that can characterize the kinetics and substrate specificity of enzymatic reactions. Nanostructured membranes can efficiently separate biomolecules4 but coupling a sensitive detection method remains difficult. Here we report a single mesoporous nano-reactor that can isolate and quantify in real-time the reaction products of proteases. The reactor consists of two layers of porous films electrochemically prepared from crystalline silicon. The upper layer with large pore sizes traps the protease enzymes and acts as the reactor while the lower layer with smaller pore sizes excludes the large proteins and captures the reaction products. Infiltration of the digested fragments into the lower layer produces a measurable change in optical reflectivity and this allows label-free quantification of enzyme kinetics in real-time within a volume of approximately 5 nanoliters. PMID:19350037

Testing piezoelectric sensors in a nuclear reactor environment

NASA Astrophysics Data System (ADS)

Reinhardt, Brian T.; Suprock, Andy; Tittmann, Bernhard

2017-02-01

Several Department of Energy Office of Nuclear Energy (DOE-NE) programs, such as the Fuel Cycle Research and Development (FCRD), Advanced Reactor Concepts (ARC), Light Water Reactor Sustainability, and Next Generation Nuclear Power Plants (NGNP), are investigating new fuels, materials, and inspection paradigms for advanced and existing reactors. A key objective of such programs is to understand the performance of these fuels and materials during irradiation. In DOE-NE's FCRD program, ultrasonic based technology was identified as a key approach that should be pursued to obtain the high-fidelity, high-accuracy data required to characterize the behavior and performance of new candidate fuels and structural materials during irradiation testing. The radiation, high temperatures, and pressure can limit the available tools and characterization methods. In this work piezoelectric transducers capable of making these measurements are developed. Specifically, three piezoelectric sensors (Bismuth Titanate, Aluminum Nitride, and Zinc Oxide) are tested in the Massachusetts Institute of Technology Research reactor to a fast neutron fluence of 8.65×1020 nf/cm2. It is demonstrated that Bismuth Titanate is capable of transduction up to 5 × 1020 nf/cm2, Zinc Oxide is capable of transduction up to at least 6.27 × 1020 nf/cm2, and Aluminum Nitride is capable of transduction up to at least 8.65 × 1020 nf/cm2.

Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger

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

Bragg-Sitton, S.M.; Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812; Kapernick, R.

2004-02-04

Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in amore » re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)« less

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

Demand driven salt clean-up in a molten salt fast reactor – Defining a priority list

PubMed Central

Litskevich, D.; Gregg, R.; Mount, A. R.

2018-01-01

The PUREX technology based on aqueous processes is currently the leading reprocessing technology in nuclear energy systems. It seems to be the most developed and established process for light water reactor fuel and the use of solid fuel. However, demand driven development of the nuclear system opens the way to liquid fuelled reactors, and disruptive technology development through the application of an integrated fuel cycle with a direct link to reactor operation. The possibilities of this new concept for innovative reprocessing technology development are analysed, the boundary conditions are discussed, and the economic as well as the neutron physical optimization parameters of the process are elucidated. Reactor physical knowledge of the influence of different elements on the neutron economy of the reactor is required. Using an innovative study approach, an element priority list for the salt clean-up is developed, which indicates that separation of Neodymium and Caesium is desirable, as they contribute almost 50% to the loss of criticality. Separating Zirconium and Samarium in addition from the fuel salt would remove nearly 80% of the loss of criticality due to fission products. The theoretical study is followed by a qualitative discussion of the different, demand driven optimization strategies which could satisfy the conflicting interests of sustainable reactor operation, efficient chemical processing for the salt clean-up, and the related economic as well as chemical engineering consequences. A new, innovative approach of balancing the throughput through salt processing based on a low number of separation process steps is developed. Next steps for the development of an economically viable salt clean-up process are identified. PMID:29494604

SAM Theory Manual

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

Hu, Rui

The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SAM development aims for advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability for reactor transient analyses. To facilitate the code development, SAM utilizes an object-oriented application framework (MOOSE), and its underlying meshing and finite-element library (libMesh) and linear and non-linear solvers (PETSc), to leverage modern advanced software environments and numerical methods. SAM focuses on modeling advanced reactormore » concepts such as SFRs (sodium fast reactors), LFRs (lead-cooled fast reactors), and FHRs (fluoride-salt-cooled high temperature reactors) or MSRs (molten salt reactors). These advanced concepts are distinguished from light-water reactors in their use of single-phase, low-pressure, high-temperature, and low Prandtl number (sodium and lead) coolants. As a new code development, the initial effort has been focused on modeling and simulation capabilities of heat transfer and single-phase fluid dynamics responses in Sodium-cooled Fast Reactor (SFR) systems. The system-level simulation capabilities of fluid flow and heat transfer in general engineering systems and typical SFRs have been verified and validated. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As the code is still under ongoing development, this SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.« less

HTR-PROTEUS pebble bed experimental program cores 9 & 10: columnar hexagonal point-on-point packing with a 1:1 moderator-to-fuel pebble ratio

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

Bess, John D.

2014-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 5, 6, 7, & 8: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:2 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

HTR-PROTEUS PEBBLE BED EXPERIMENTAL PROGRAM CORES 9 & 10: COLUMNAR HEXAGONAL POINT-ON-POINT PACKING WITH A 1:1 MODERATOR-TO-FUEL PEBBLE RATIO

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

John D. Bess

2013-03-01

PROTEUS is a zero-power research reactor based on a cylindrical graphite annulus with a central cylindrical cavity. The graphite annulus remains basically the same for all experimental programs, but the contents of the central cavity are changed according to the type of reactor being investigated. Through most of its service history, PROTEUS has represented light-water reactors, but from 1992 to 1996 PROTEUS was configured as a pebble-bed reactor (PBR) critical facility and designated as HTR-PROTEUS. The nomenclature was used to indicate that this series consisted of High Temperature Reactor experiments performed in the PROTEUS assembly. During this period, seventeen criticalmore » configurations were assembled and various reactor physics experiments were conducted. These experiments included measurements of criticality, differential and integral control rod and safety rod worths, kinetics, reaction rates, water ingress effects, and small sample reactivity effects (Ref. 3). HTR-PROTEUS was constructed, and the experimental program was conducted, for the purpose of providing experimental benchmark data for assessment of reactor physics computer codes. Considerable effort was devoted to benchmark calculations as a part of the HTR-PROTEUS program. References 1 and 2 provide detailed data for use in constructing models for codes to be assessed. Reference 3 is a comprehensive summary of the HTR-PROTEUS experiments and the associated benchmark program. This document draws freely from these references. Only Cores 9 and 10 are evaluated in this benchmark report due to similarities in their construction. The other core configurations of the HTR-PROTEUS program are evaluated in their respective reports as outlined in Section 1.0. Cores 9 and 10 were evaluated and determined to be acceptable benchmark experiments.« less

Nuclear fuel management optimization using genetic algorithms

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

DeChaine, M.D.; Feltus, M.A.

1995-07-01

The code independent genetic algorithm reactor optimization (CIGARO) system has been developed to optimize nuclear reactor loading patterns. It uses genetic algorithms (GAs) and a code-independent interface, so any reactor physics code (e.g., CASMO-3/SIMULATE-3) can be used to evaluate the loading patterns. The system is compared to other GA-based loading pattern optimizers. Tests were carried out to maximize the beginning of cycle k{sub eff} for a pressurized water reactor core loading with a penalty function to limit power peaking. The CIGARO system performed well, increasing the k{sub eff} after lowering the peak power. Tests of a prototype parallel evaluation methodmore » showed the potential for a significant speedup.« less

Latest progress from the Daya Bay reactor neutrino experiment

NASA Astrophysics Data System (ADS)

Wang, Zhe; Daya Bay Collaboration

2016-05-01

Recently the Daya Bay reactor neutrino experiment has presented several new results about neutrino and reactor physics after acquiring a large data sample and after gaining a more sophisticated understanding of the experiment. In this talk I will introduce the latest progress made by the experiment including a three-flavor neutrino oscillation analysis using neutron capture on gadolinium, which gave sin2 2θ 13 = 0.084 ± 0.005 and |Δm2 ee| = (2.42 ±0.11) × 10-3 eV2, an independent θ 13 measurement using neutron capture on hydrogen, a search for a light sterile neutrino, and a measurement of the reactor antineutrino flux and spectrum.

Development of the reactor antineutrino detection technology within the iDream project

NASA Astrophysics Data System (ADS)

Gromov, M.; Kuznetsov, D.; Murchenko, A.; Novikova, G.; Obinyakov, B.; Oralbaev, A.; Plakitina, K.; Skorokhvatov, M.; Sukhotin, S.; Chepurnov, A.; Etenko, A.

2017-12-01

The iDREAM (industrial Detector for reactor antineutrino monitoring) project is aimed at remote monitoring of the operating modes of the atomic reactor on nuclear power plant to ensure a technical support of IAEA non-proliferation safeguards. The detector is a scintillator spectrometer. The sensitive volume (target) is filled with a liquid organic scintillator based on linear alkylbenzene where reactor antineutrinos will be detected via inverse beta-decay reaction. We present first results of laboratory tests after physical launch. The detector was deployed at sea level without background shielding. The number of calibrations with radioactive sources was conducted. All data were obtained by means of a slow control system which was put into operation.

A new dual-collimation batch reactor for determination of ultraviolet inactivation rate constants for microorganisms in aqueous suspensions

PubMed Central

Martin, Stephen B.; Schauer, Elizabeth S.; Blum, David H.; Kremer, Paul A.; Bahnfleth, William P.; Freihaut, James D.

2017-01-01

We developed, characterized, and tested a new dual-collimation aqueous UV reactor to improve the accuracy and consistency of aqueous k-value determinations. This new system is unique because it collimates UV energy from a single lamp in two opposite directions. The design provides two distinct advantages over traditional single-collimation systems: 1) real-time UV dose (fluence) determination; and 2) simple actinometric determination of a reactor factor that relates measured irradiance levels to actual irradiance levels experienced by the microbial suspension. This reactor factor replaces three of the four typical correction factors required for single-collimation reactors. Using this dual-collimation reactor, Bacillus subtilis spores demonstrated inactivation following the classic multi-hit model with k = 0.1471 cm2/mJ (with 95% confidence bounds of 0.1426 to 0.1516). PMID:27498232

Statistical Exposé of a Multiple-Compartment Anaerobic Reactor Treating Domestic Wastewater.

PubMed

Pfluger, Andrew R; Hahn, Martha J; Hering, Amanda S; Munakata-Marr, Junko; Figueroa, Linda

2018-06-01

  Mainstream anaerobic treatment of domestic wastewater is a promising energy-generating treatment strategy; however, such reactors operated in colder regions are not well characterized. Performance data from a pilot-scale, multiple-compartment anaerobic reactor taken over 786 days were subjected to comprehensive statistical analyses. Results suggest that chemical oxygen demand (COD) was a poor proxy for organics in anaerobic systems as oxygen demand from dissolved inorganic material, dissolved methane, and colloidal material influence dissolved and particulate COD measurements. Additionally, univariate and functional boxplots were useful in visualizing variability in contaminant concentrations and identifying statistical outliers. Further, significantly different dissolved organic removal and methane production was observed between operational years, suggesting that anaerobic reactor systems may not achieve steady-state performance within one year. Last, modeling multiple-compartment reactor systems will require data collected over at least two years to capture seasonal variations of the major anaerobic microbial functions occurring within each reactor compartment.

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.

NGNP Data Management and Analysis System Analysis and Web Delivery Capabilities

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

Cynthia D. Gentillon

2010-09-01

Projects for the Very High Temperature Reactor Technology Development Office provide data in support of Nuclear Regulatory Commission licensing of the very high temperature reactor. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high-temperature and high-fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The Very High Temperature Reactor Technology Development Office has established the NGNP Data Management and Analysis System (NDMAS) at the Idaho National Laboratory to ensure that very high temperature reactor data are (1) qualified for use, (2) stored in amore » readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the third NDMAS objective. It describes capabilities for displaying the data in meaningful ways and for data analysis to identify useful relationships among the measured quantities.« less

A feasibility assessment of installation, operation and disposal options for nuclear reactor power system concepts for a NASA growth space station

NASA Technical Reports Server (NTRS)

Bloomfield, Harvey S.; Heller, Jack A.

1987-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth space station architecture was conducted to address a variety of installation, operational disposition, and safety issues. A previous NASA sponsored study, which showed the advantages of space station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide the feasibility of each combination.

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

A feasibility assessment of nuclear reactor power system concepts for the NASA Growth Space Station

NASA Technical Reports Server (NTRS)

Bloomfield, H. S.; Heller, J. A.

1986-01-01

A preliminary feasibility assessment of the integration of reactor power system concepts with a projected growth Space Station architecture was conducted to address a variety of installation, operational, disposition and safety issues. A previous NASA sponsored study, which showed the advantages of Space Station - attached concepts, served as the basis for this study. A study methodology was defined and implemented to assess compatible combinations of reactor power installation concepts, disposal destinations, and propulsion methods. Three installation concepts that met a set of integration criteria were characterized from a configuration and operational viewpoint, with end-of-life disposal mass identified. Disposal destinations that met current aerospace nuclear safety criteria were identified and characterized from an operational and energy requirements viewpoint, with delta-V energy requirement as a key parameter. Chemical propulsion methods that met current and near-term application criteria were identified and payload mass and delta-V capabilities were characterized. These capabilities were matched against concept disposal mass and destination delta-V requirements to provide a feasibility of each combination.

The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

NASA Astrophysics Data System (ADS)

Dautray, Robert

2011-06-01

The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation, etc., radioprotection measures and treatment for the "transuranic" elements. For a long period of time, France was in the forefront of nuclear breeder power generation science, technological research and also in the knowledge base related to breeder reactors. It is in the country's interest to pursue these efforts and this could per se constitute one of the national priorities. Nous sommes naturellement bien conscients de l'énorme problème qui se pose au Japon actuellement comme suite au tremblement de terre et au tsunami de mars 2011 et leurs conséquences, notamment sur des installations électronucléaires. Le texte que nous présentons concerne des conditions totalement générales, indépendantes des problèmes spécifiques de sûreté qu'il faudra, de toute façon, traiter dans le cadre d'un développement éventuel de l'énergie nucléaire.We are aware, of course, of the huge problem that Japan has to deal with the aftermath of the quake and tsunami of March 2011 and their consequences on electronuclear power plants. The text that we present here concerns general physical topics independent of the specific safety problems, general physical topics which will have to be solved in the case of a contingent development of electronuclear power plants.

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

Gorelenkov, Nikolai N

The area of energetic particle (EP) physics of fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by W.W. Heidbrink and G.J. Sadler [1]. That review coincided with the start of deuterium-tritium (DT) experiments on Tokamak Fusion Test reactor (TFTR) and full scale fusion alphas physics studies. Fusion research in recent years has been influenced by EP physics in many ways including the limitations imposed by the "sea" of Alfven eigenmodes (AE) in particular by the toroidicityinduced AEsmore » (TAE) modes and reversed shear Alfven (RSAE). In present paper we attempt a broad review of EP physics progress in tokamaks and spherical tori since the first DT experiments on TFTR and JET (Joint European Torus) including helical/stellarator devices. Introductory discussions on basic ingredients of EP physics, i.e. particle orbits in STs, fundamental diagnostic techniques of EPs and instabilities, wave particle resonances and others are given to help understanding the advanced topics of EP physics. At the end we cover important and interesting physics issues toward the burning plasma experiments such as ITER (International Thermonuclear Experimental Reactor).« less

Nuclear reactor for breeding U.sup.233

DOEpatents

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

1976-01-01

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

Planetary surface reactor shielding using indigenous materials

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

Houts, Michael G.; Poston, David I.; Trellue, Holly R.

The exploration and development of Mars will require abundant surface power. Nuclear reactors are a low-cost, low-mass means of providing that power. A significant fraction of the nuclear power system mass is radiation shielding necessary for protecting humans and/or equipment from radiation emitted by the reactor. For planetary surface missions, it may be desirable to provide some or all of the required shielding from indigenous materials. This paper examines shielding options that utilize either purely indigenous materials or a combination of indigenous and nonindigenous materials. {copyright} {ital 1999 American Institute of Physics.}

THE ARMOUR DUST FUELED REACTOR (ADFR). Quarterly Progress Report No. 1 for the Period February 21, 1958 to May 21, 1958

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

Loewe, W.E.; Krucoff, D.

1958-10-31

Work has begun on the ADFR, a reactor using a new fuel form -- fissionable dust carried in an inent gas. Temperatures in the range 2,000 to 3,000 deg F appear feasible in an all-ceramic system. Experimental study of the fuel form was initiated, and a loop to circulate the fuel dust was constructed. Initial operation is encouraging. Theoretical studies were carried on in the areas of reactor physics, heat transfer, and safety. (auth)

Elementary Reactions and Their Role in Gas-Phase Prebiotic Chemistry

PubMed Central

Balucani, Nadia

2009-01-01

The formation of complex organic molecules in a reactor filled with gaseous mixtures possibly reproducing the primitive terrestrial atmosphere and ocean demonstrated more than 50 years ago that inorganic synthesis of prebiotic molecules is possible, provided that some form of energy is provided to the system. After that groundbreaking experiment, gas-phase prebiotic molecules have been observed in a wide variety of extraterrestrial objects (including interstellar clouds, comets and planetary atmospheres) where the physical conditions vary widely. A thorough characterization of the chemical evolution of those objects relies on a multi-disciplinary approach: 1) observations allow us to identify the molecules and their number densities as they are nowadays; 2) the chemistry which lies behind their formation starting from atoms and simple molecules is accounted for by complex reaction networks; 3) for a realistic modeling of such networks, a number of experimental parameters are needed and, therefore, the relevant molecular processes should be fully characterized in laboratory experiments. A survey of the available literature reveals, however, that much information is still lacking if it is true that only a small percentage of the elementary reactions considered in the models have been characterized in laboratory experiments. New experimental approaches to characterize the relevant elementary reactions in laboratory are presented and the implications of the results are discussed. PMID:19564951

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.

Progress In Developing Laser Based Post Irradiation Examination Infrastructure

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

Smith, James A.; Scott, Clark L.; Benefiel, Brad C.

To be able to understand the performance of reactor fuels and materials, irradiated materials must be characterized effectively and efficiently in a high rad environment. The characterization work must be performed remotely and in an environment hostile to instrumentation. Laser based characterization techniques provide the ability to be remote and robust in a hot-cell environment. Laser based instrumentation also can provide high spatial resolution suitable for scanning and imaging large areas. The INL is currently developing three laser based Post Irradiation Examination (PIE) stations for the Hot Fuel Examination Facility at the INL. These laser based systems will characterize irradiatedmore » materials and fuels. The characterization systems are the following: Laser Shock Laser based ultrasonic C-scan system Gas Assay, Sample, and Recharge system (GASR, up-grade to an existing system). The laser shock technique will characterize material properties and failure loads/mechanisms in various materials such as LWR fuel, plate fuel, and next generation fuel forms, for PIE in high radiation areas. The laser shock-technique induces large amplitude shock waves to mechanically characterize interfaces such as the fuel-clad bond. The shock wave travels as a compression wave through the material to the free (unconfined) back surface and reflects back through the material under test as a rarefaction (tensile) wave. This rarefaction wave is the physical mechanism that produces internal de-lamination failure. As part of the laser shock system, a laser-based ultrasonic C-scan system will be used to detect and characterize debonding caused by the laser shock technique. The laser ultrasonic system will be fully capable of performing classical non-destructive evaluation testing and imaging functions such as microstructure characterization, flaw detection and dimensional metrology in complex components. The purpose of the GASR is to measure the pressure/volume of the plenum of an irradiated fuel element and obtain fission gas samples for analysis. The study of pressure and volume in the plenum of an irradiated fuel element and the analysis of fission gases released from the fuel is important to understanding the performance of reactor fuels and materials. This system may also be used to measure the pressure/volume of other components (such as control blades) and obtain gas samples from these components for analysis. The main function of the laser in this application is to puncture the fuel element to allow the fission gas to escape and if necessary to weld the spot close. The GASR station will have the inherent capability to perform cutting welding and joining functions within a hot-cell.« less

Nuclear Data Needs for Generation IV Nuclear Energy Systems

NASA Astrophysics Data System (ADS)

Rullhusen, Peter

2006-04-01

Nuclear data needs for generation IV systems. Future of nuclear energy and the role of nuclear data / P. Finck. Nuclear data needs for generation IV nuclear energy systems-summary of U.S. workshop / T. A. Taiwo, H. S. Khalil. Nuclear data needs for the assessment of gen. IV systems / G. Rimpault. Nuclear data needs for generation IV-lessons from benchmarks / S. C. van der Marck, A. Hogenbirk, M. C. Duijvestijn. Core design issues of the supercritical water fast reactor / M. Mori ... [et al.]. GFR core neutronics studies at CEA / J. C. Bosq ... [et al]. Comparative study on different phonon frequency spectra of graphite in GCR / Young-Sik Cho ... [et al.]. Innovative fuel types for minor actinides transmutation / D. Haas, A. Fernandez, J. Somers. The importance of nuclear data in modeling and designing generation IV fast reactors / K. D. Weaver. The GIF and Mexico-"everything is possible" / C. Arrenondo Sánchez -- Benmarks, sensitivity calculations, uncertainties. Sensitivity of advanced reactor and fuel cycle performance parameters to nuclear data uncertainties / G. Aliberti ... [et al.]. Sensitivity and uncertainty study for thermal molten salt reactors / A. Biduad ... [et al.]. Integral reactor physics benchmarks- The International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPHEP) / J. B. Briggs, D. W. Nigg, E. Sartori. Computer model of an error propagation through micro-campaign of fast neutron gas cooled nuclear reactor / E. Ivanov. Combining differential and integral experiments on [symbol] for reducing uncertainties in nuclear data applications / T. Kawano ... [et al.]. Sensitivity of activation cross sections of the Hafnium, Tanatalum and Tungsten stable isotopes to nuclear reaction mechanisms / V. Avrigeanu ... [et al.]. Generating covariance data with nuclear models / A. J. Koning. Sensitivity of Candu-SCWR reactors physics calculations to nuclear data files / K. S. Kozier, G. R. Dyck. The lead cooled fast reactor benchmark BREST-300: analysis with sensitivity method / V. Smirnov ... [et al.]. Sensitivity analysis of neutron cross-sections considered for design and safety studies of LFR and SFR generation IV systems / K. Tucek, J. Carlsson, H. Wider -- Experiments. INL capabilities for nuclear data measurements using the Argonne intense pulsed neutron source facility / J. D. Cole ... [et al.]. Cross-section measurements in the fast neutron energy range / A. Plompen. Recent measurements of neutron capture cross sections for minor actinides by a JNC and Kyoto University Group / H. Harada ... [et al.]. Determination of minor actinides fission cross sections by means of transfer reactions / M. Aiche ... [et al.] -- Evaluated data libraries. Nuclear data services from the NEA / H. Henriksson, Y. Rugama. Nuclear databases for energy applications: an IAEA perspective / R. Capote Noy, A. L. Nichols, A. Trkov. Nuclear data evaluation for generation IV / G. Noguère ... [et al.]. Improved evaluations of neutron-induced reactions on americium isotopes / P. Talou ... [et al.]. Using improved ENDF-based nuclear data for candu reactor calculations / J. Prodea. A comparative study on the graphite-moderated reactors using different evaluated nuclear data / Do Heon Kim ... [et al.].

Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

PubMed

Hill, R N; Nutt, W M; Laidler, J J

2011-01-01

The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described. Copyright © 2010 Health Physics Society

Modelling the attenuation in the ATHENA finite elements code for the ultrasonic testing of austenitic stainless steel welds.

PubMed

Chassignole, B; Duwig, V; Ploix, M-A; Guy, P; El Guerjouma, R

2009-12-01

Multipass welds made in austenitic stainless steel, in the primary circuit of nuclear power plants with pressurized water reactors, are characterized by an anisotropic and heterogeneous structure that disturbs the ultrasonic propagation and makes ultrasonic non-destructive testing difficult. The ATHENA 2D finite element simulation code was developed to help understand the various physical phenomena at play. In this paper, we shall describe the attenuation model implemented in this code to give an account of wave scattering phenomenon through polycrystalline materials. This model is in particular based on the optimization of two tensors that characterize this material on the basis of experimental values of ultrasonic velocities attenuation coefficients. Three experimental configurations, two of which are representative of the industrial welds assessment case, are studied in view of validating the model through comparison with the simulation results. We shall thus provide a quantitative proof that taking into account the attenuation in the ATHENA code dramatically improves the results in terms of the amplitude of the echoes. The association of the code and detailed characterization of a weld's structure constitutes a remarkable breakthrough in the interpretation of the ultrasonic testing on this type of component.

Strontium-90 adsorption-desorption properties and sediment characterization at the 100 N-Area

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

Serne, R.J.; LeGore, V.L.

1996-01-01

Strontium-90 ({sup 90}Sr) has been seeping into the Columbia River since the early 1980s. The likely source is subsurface migration of {sup 90}Sr from once-through cooling water from the Hanford N Reactor disposed into the two disposal crib/trench facilities. Background information has been provided on the operational history of the two liquid waste disposal facilities and some of the regulatory drivers that have lead to the various characterization activities and remediation demonstrations being performed to help choose future full-scale remediation alternatives. The work presented in this topical report had two main objectives. First, we obtained numerous borehole samples from newlymore » installed wells/borings and performed physical and chemical characterization that,included particle size analysis, moisture content, and Strontium-90, Tritium and gamma activity analyses to help improve the conceptual model of where the contaminants currently reside in the sediments. The second objective was to perform laboratory adsorption-desorption tests using both batch and flow- through column techniques to gather data for use in contaminant transport conceptual models and to aid in specific pump-and-treat calculations needed to interpret a field demonstration.« less

Environmental characterization of two potential locations at Hanford for a new production reactor

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

Watson, E.C.; Becker, C.D.; Fitzner, R.E.

This report describes various environmental aspects of two areas on the Hanford Site that are potential locations for a New Production Reactor (NPR). The area known as the Skagit Hanford Site is considered the primary or reference site. The second area, termed the Firehouse Site, is considered the alternate site. The report encompasses an environmental characterization of these two potential NPR locations. Eight subject areas are covered: geography and demography; ecology; meteorology; hydrology; geology; cultural resources assessment; economic and social effects of station construction and operation; and environmental monitoring. 80 refs., 68 figs., 109 tabs.

Characterization of Filters Loaded With Reactor Strontium Carbonate - 13203

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

Josephson, Walter S.; Steen, Franciska H.

A collection of three highly radioactive filters containing reactor strontium carbonate were being prepared for disposal. All three filters were approximately characterized at the time of manufacture by gravimetric methods. The first filter had been partially emptied, and the quantity of residual activity was uncertain. Dose rate to activity modeling using the Monte-Carlo N Particle (MCNP) code was selected to confirm the gravimetric characterization of the full filters, and to fully characterize the partially emptied filter. Although dose rate to activity modeling using MCNP is a common technique, it is not often used for Bremsstrahlung-dominant materials such as reactor strontium.more » As a result, different MCNP modeling options were compared to determine the optimum approach. This comparison indicated that the accuracy of the results were heavily dependent on the MCNP modeling details and the location of the dose rate measurement point. The optimum model utilized a photon spectrum generated by the Oak Ridge Isotope Generation and Depletion (ORIGEN) code and dose rates measured at 30 cm. Results from the optimum model agreed with the gravimetric estimates within 15%. It was demonstrated that dose rate to activity modeling can be successful for Bremsstrahlung-dominant radioactive materials. However, the degree of success is heavily dependent on the choice of modeling techniques. (authors)« less

Lawrence Livermore National Laboratory and Sandia National Laboratory Nuclear Accident Dosimetry Support of IER 252 and the Dose Characterization of the Flattop Reactor at the DAF

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

Hickman, D. P.; Jeffers, K. L.; Radev, R. P.

In support of IER 252 “Characterization of the Flattop Reactor at the NCERC”, LLNL performed ROSPEC measurements of the neutron spectrum and deployed 129 Personnel Nuclear Accident Dosimeters (PNAD) to establish the need for height corrections and verification of neutron spectrum evaluation of the fluences and dose. A very limited number of heights (typically only one or two heights) can be measured using neutron spectrometers, therefore it was important to determine if any height correction would be needed in future intercomparisons and studies. Specific measurement positions around the Flatttop reactor are provided in Figure 1. Table 1 provides run andmore » position information for LLNL measurements. The LLNL ROSPEC (R2) was used for run numbers 1 – 7, and vi. PNADs were positioned on trees during run numbers 9, 11, and 13.« less

Design and installation of a ferromagnetic wall in tokamak geometry

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

Hughes, P. E., E-mail: peh2109@columbia.edu; Levesque, J. P.; Rivera, N.

Low-activation ferritic steels are leading material candidates for use in next-generation fusion development experiments such as a prospective component test facility and DEMO power reactor. Understanding the interaction of plasmas with a ferromagnetic wall will provide crucial physics for these facilities. In order to study ferromagnetic effects in toroidal geometry, a ferritic wall upgrade was designed and installed in the High Beta Tokamak–Extended Pulse (HBT-EP). Several material options were investigated based on conductivity, magnetic permeability, vacuum compatibility, and other criteria, and the material of choice (high-cobalt steel) is characterized. Installation was accomplished quickly, with minimal impact on existing diagnostics andmore » overall machine performance, and initial results demonstrate the effects of the ferritic wall on plasma stability.« less

Brookhaven highlights, October 1979-September 1980

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

Not Available

1980-01-01

Highlights are given for the research areas of the Brookhaven National Laboratory. These areas include high energy physics, physics and chemistry, life sciences, applied energy science (energy and environment, and nuclear energy), and support activities (including mathematics, instrumentation, reactors, and safety). (GHT)

Coupling Schemes for Multiphysics Reactor Simulation

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

Vijay Mahadeven; Jean Ragusa

2007-11-01

This report documents the progress of the student Vijay S. Mahadevan from the Nuclear Engineering Department of Texas A&M University over the summer of 2007 during his visit to the INL. The purpose of his visit was to investigate the physics-based preconditioned Jacobian-free Newton-Krylov method applied to physics relevant to nuclear reactor simulation. To this end he studied two test problems that represented reaction-diffusion and advection-reaction. These two test problems will provide the basis for future work in which neutron diffusion, nonlinear heat conduction, and a twophase flow model will be tightly coupled to provide an accurate model of amore » BWR core.« less

MTR WING, TRA604. FIRST FLOOR PLAN. ENTRY LOBBY, MACHINE SHOP, ...

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

MTR WING, TRA-604. FIRST FLOOR PLAN. ENTRY LOBBY, MACHINE SHOP, INSTRUMENT SHOP, COUNTING ROOM, HEALTH PHYSICS LAB, LABS AND OFFICES, STORAGE, SHIPPING AND RECEIVING. BLAW-KNOX 3150-4-2, 7/1950. INL INDEX NO. 053-604-00-099-100008, REV. 7. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

ERIC Educational Resources Information Center

Fillo, J. A.

This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

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

NONE

During this reporting period, there were three major thrusts in the WVU portion. First, we started a preliminary investigation on the use of a membrane reactor for HAS. Accordingly, the plug-flow reactor which had been isolated from sulfides was substituted by a membrane reactor. The tubular membrane was first characterized in terms of its permeation properties, i.e., the fluxes, permeances and selectivities of the components. After that, a BASF methanol-synthesis catalyst was tested under different conditions on the membrane reactor. The results will be compared with those from a non-permeable stainless steel tubular reactor under the same conditions. Second, wemore » started a detailed study of one of the catalysts tested during the screening runs. Accordingly, a carbon-supported potassium-doped molybdenum-cobalt catalyst was selected to be run in the Rotoberty reactor. Finally, we have started detailed analyses of reaction products from some earlier screening runs in which non-sulfide molybdenum-based catalysts were employed and much more complicated product distributions were generally observed. These products could not hitherto be analyzed using the gas chromatograph which was then available. A Varian gas chromatograph/mass spectrometer (GC/MS) is being used to characterize these liquid products. At UCC, we completed a screening of an Engelhard support impregnated with copper and cesium. We have met or exceeded three of four catalyst development targets. Oxygenate selectivity is our main hurdle. Further, we tested the effect of replacing stainless-steel reactor preheater tubing and fittings with titanium ones. We had hoped to reduce the yield of hydrocarbons which may have been produced at high temperatures due to Fischer-Tropsch catalysis with the iron and nickel in the preheater tube walls. Results showed that total hydrocarbon space time yield was actually increased with the titanium preheater, while total alcohol space time yield was not significantly affected.« less

Predictive characterization of aging and degradation of reactor materials in extreme environments. Final report, December 20, 2013 - September 20, 2017

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

Qu, Jianmin

Understanding of reactor material behavior in extreme environments is vital not only to the development of new materials for the next generation nuclear reactors, but also to the extension of the operating lifetimes of the current fleet of nuclear reactors. To this end, this project conducted a suite of unique experimental techniques, augmented by a mesoscale computational framework, to understand and predict the long-term effects of irradiation, temperature, and stress on material microstructures and their macroscopic behavior. The experimental techniques and computational tools were demonstrated on two distinctive types of reactor materials, namely, Zr alloys and high-Cr martensitic steels. Thesemore » materials are chosen as the test beds because they are the archetypes of high-performance reactor materials (cladding, wrappers, ducts, pressure vessel, piping, etc.). To fill the knowledge gaps, and to meet the technology needs, a suite of innovative in situ transmission electron microscopy (TEM) characterization techniques (heating, heavy ion irradiation, He implantation, quantitative small-scale mechanical testing, and various combinations thereof) were developed and used to elucidate and map the fundamental mechanisms of microstructure evolution in both Zr and Cr alloys for a wide range environmental boundary conditions in the thermal-mechanical-irradiation input space. Knowledge gained from the experimental observations of the active mechanisms and the role of local microstructural defects on the response of the material has been incorporated into a mathematically rigorous and comprehensive three-dimensional mesoscale framework capable of accounting for the compositional variation, microstructural evolution and localized deformation (radiation damage) to predict aging and degradation of key reactor materials operating in extreme environments. Predictions from this mesoscale framework were compared with the in situ TEM observations to validate the model.« less

SCALE Code System

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

Jessee, Matthew Anderson

The SCALE Code System is a widely-used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including three deterministicmore » and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.SCALE 6.2 provides many new capabilities and significant improvements of existing features.New capabilities include:• ENDF/B-VII.1 nuclear data libraries CE and MG with enhanced group structures,• Neutron covariance data based on ENDF/B-VII.1 and supplemented with ORNL data,• Covariance data for fission product yields and decay constants,• Stochastic uncertainty and correlation quantification for any SCALE sequence with Sampler,• Parallel calculations with KENO,• Problem-dependent temperature corrections for CE calculations,• CE shielding and criticality accident alarm system analysis with MAVRIC,• CE depletion with TRITON (T5-DEPL/T6-DEPL),• CE sensitivity/uncertainty analysis with TSUNAMI-3D,• Simplified and efficient LWR lattice physics with Polaris,• Large scale detailed spent fuel characterization with ORIGAMI and ORIGAMI Automator,• Advanced fission source convergence acceleration capabilities with Sourcerer,• Nuclear data library generation with AMPX, and• Integrated user interface with Fulcrum.Enhanced capabilities include:• Accurate and efficient CE Monte Carlo methods for eigenvalue and fixed source calculations,• Improved MG resonance self-shielding methodologies and data,• Resonance self-shielding with modernized and efficient XSProc integrated into most sequences,• Accelerated calculations with TRITON/NEWT (generally 4x faster than SCALE 6.1),• Spent fuel characterization with 1470 new reactor-specific libraries for ORIGEN,• Modernization of ORIGEN (Chebyshev Rational Approximation Method [CRAM] solver, API for high-performance depletion, new keyword input format)• Extension of the maximum mixture number to values well beyond the previous limit of 2147 to ~2 billion,• Nuclear data formats enabling the use of more than 999 energy groups,• Updated standard composition library to provide more accurate use of natural abundances, andvi• Numerous other enhancements for improved usability and stability.« less

IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis

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

Strydom, Gerhard; Bostelmann, F.

The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics and depletion simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, modelling and computational algorithms. (The interested reader is referred to the large body of published SA and UA literature for a more complete overview of the various types of uncertainties, methodologies and results obtained).more » SA is helpful for ranking the various sources of uncertainty and error in the results of core analyses. SA and UA are required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. SA and UA can guide experimental, modelling, and algorithm research and development. Current SA and UA rely either on derivative-based methods such as stochastic sampling methods or on generalized perturbation theory to obtain sensitivity coefficients. Neither approach addresses all needs. In order to benefit from recent advances in modelling and simulation and the availability of new covariance data (nuclear data uncertainties) extensive sensitivity and uncertainty studies are needed for quantification of the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more robust and well validated calculation tools to meet design target accuracies. In February 2009, the Technical Working Group on Gas-Cooled Reactors (TWG-GCR) of the International Atomic Energy Agency (IAEA) recommended that the proposed Coordinated Research Program (CRP) on the HTGR Uncertainty Analysis in Modelling (UAM) be implemented. This CRP is a continuation of the previous IAEA and Organization for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) international activities on Verification and Validation (V&V) of available analytical capabilities for HTGR simulation for design and safety evaluations. Within the framework of these activities different numerical and experimental benchmark problems were performed and insight was gained about specific physics phenomena and the adequacy of analysis methods.« less

Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

DOE PAGES

Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

2015-11-04

We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO 2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize themore » degradation rate constant.« less

An Optically Accessible Pyrolysis Microreactor

NASA Astrophysics Data System (ADS)

Baraban, Joshua H.; David, Donald E.; Ellison, Barney; Daily, John W.

2016-06-01

We report an optically accessible pyrolysis micro-reactor suitable for in situ laser spectroscopic measurements. A radiative heating design allows for completely unobstructed views of the micro-reactor along two axes. The maximum temperature demonstrated here is only 1300 K (as opposed to 1700 K for the usual SiC micro-reactor) because of the melting point of fused silica, but alternative transparent materials will allow for higher temperatures. Laser induced fluorescence measurements on nitric oxide are presented as a proof of principle for spectroscopic characterization of pyrolysis conditions. (This work has been published in J. H. Baraban, D. E. David, G. B. Ellison, and J. W. Daily. An Optically Accessible Pyrolysis Micro-Reactor. Review of Scientific Instruments, 87(1):014101, 2016.)

Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

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

Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO 2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize themore » degradation rate constant.« less

Overview of past and current activities on fuels for fast reactors at the Institute for Transuranium Elements

NASA Astrophysics Data System (ADS)

Fernandez, A.; McGinley, J.; Somers, J.; Walter, M.

2009-07-01

Nuclear energy has the potential to provide a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gas emissions. The renewal of interest in fast neutron spectra reactors to meet more ambitious sustainable development criteria (i.e., resource maximisation and waste minimisation), opens a favourable framework for R&D activities in this area. The Institute for Transuranium Elements has extensive experience in the fabrication, characterization and irradiation testing (Phénix, Dounreay, Rapsodie) of fast reactor fuels, in oxide, nitride and carbide forms. An overview of these past and current activities on fast reactor fuels is presented.

Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

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

Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.

A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

The CANDU Reactor System: An Appropriate Technology.

PubMed

Robertson, J A

1978-02-10

CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity.

Supply of enriched uranium for research reactors

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

Mueller, H.

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel onmore » December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.« less

Modeling residence-time distribution in horizontal screw hydrolysis reactors

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

Sievers, David A.; Stickel, Jonathan J.

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Current and future trends for biofilm reactors for fermentation processes.

PubMed

Ercan, Duygu; Demirci, Ali

2015-03-01

Biofilms in the environment can both cause detrimental and beneficial effects. However, their use in bioreactors provides many advantages including lesser tendencies to develop membrane fouling and lower required capital costs, their higher biomass density and operation stability, contribution to resistance of microorganisms, etc. Biofilm formation occurs naturally by the attachment of microbial cells to the support without use of any chemicals agent in biofilm reactors. Biofilm reactors have been studied and commercially used for waste water treatment and bench and pilot-scale production of value-added products in the past decades. It is important to understand the fundamentals of biofilm formation, physical and chemical properties of a biofilm matrix to run the biofilm reactor at optimum conditions. This review includes the principles of biofilm formation; properties of a biofilm matrix and their roles in the biofilm formation; factors that improve the biofilm formation, such as support materials; advantages and disadvantages of biofilm reactors; and industrial applications of biofilm reactors.

Modeling residence-time distribution in horizontal screw hydrolysis reactors

DOE PAGES

Sievers, David A.; Stickel, Jonathan J.

2017-10-12

The dilute-acid thermochemical hydrolysis step used in the production of liquid fuels from lignocellulosic biomass requires precise residence-time control to achieve high monomeric sugar yields. Difficulty has been encountered reproducing residence times and yields when small batch reaction conditions are scaled up to larger pilot-scale horizontal auger-tube type continuous reactors. A commonly used naive model estimated residence times of 6.2-16.7 min, but measured mean times were actually 1.4-2.2 the estimates. Here, this study investigated how reactor residence-time distribution (RTD) is affected by reactor characteristics and operational conditions, and developed a method to accurately predict the RTD based on key parameters.more » Screw speed, reactor physical dimensions, throughput rate, and process material density were identified as major factors affecting both the mean and standard deviation of RTDs. The general shape of RTDs was consistent with a constant value determined for skewness. The Peclet number quantified reactor plug-flow performance, which ranged between 20 and 357.« less

Investigation of materials for fusion power reactors

NASA Astrophysics Data System (ADS)

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

2014-06-01

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

Solar gasification of biomass: design and characterization of a molten salt gasification reactor

NASA Astrophysics Data System (ADS)

Hathaway, Brandon Jay

The design and implementation of a prototype molten salt solar reactor for gasification of biomass is a significant milestone in the development of a solar gasification process. The reactor developed in this work allows for 3 kWth operation with an average aperture flux of 1530 suns at salt temperatures of 1200 K with pneumatic injection of ground or powdered dry biomass feedstocks directly into the salt melt. Laboratory scale experiments in an electrically heated reactor demonstrate the benefits of molten salt and the data was evaluated to determine the kinetics of pyrolysis and gasification of biomass or carbon in molten salt. In the presence of molten salt overall gas yields are increased by up to 22%; pyrolysis rates double due to improved heat transfer, while carbon gasification rates increase by an order of magnitude. Existing kinetic models for cellulose pyrolysis fit the data well, while carbon gasification in molten salt follows kinetics modeled with a 2/3 order shrinking-grain model with a pre-exponential factor of 1.5*106 min-1 and activation energy of 158 kJ/mol. A reactor concept is developed based around a concentric cylinder geometry with a cavity-style solar receiver immersed within a volume of molten carbonate salt. Concentrated radiation delivered to the cavity is absorbed in the cavity walls and transferred via convection to the salt volume. Feedstock is delivered into the molten salt volume where biomass gasification reactions will be carried out producing the desired product gas. The features of the cavity receiver/reactor concept are optimized based on modeling of the key physical processes. The cavity absorber geometry is optimized according to a parametric survey of radiative exchange using a Monte Carlo ray tracing model, resulting in a cavity design that achieves absorption efficiencies of 80%-90%. A parametric survey coupling the radiative exchange simulations to a CFD model of molten salt natural convection is used to size the annulus containing the molten salt to maximize utilization of absorbed solar energy, resulting in a predicted utilization efficiency of 70%. Finite element analysis was used to finalize the design to achieve acceptable thermal stresses less than 34.5 MPa to avoid material creep.

Multi-physics design and analyses of long life reactors for lunar outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.

Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete Element Method (DEM) analysis in lunar gravity. In addition, this research addresses the post-operation storage of the SCoRe and PeBR concepts, below the lunar surface, to determine the time required for the radioactivity in the used fuel to decrease to a low level to allow for its safe recovery. The SCoRe and PeBR concepts are designed to operate at coolant temperatures ≤ 900 K and use conventional stainless steels and superalloys for the structure in the reactor core and power system. They are emplaced below grade on the Moon to take advantage of the regolith as a supplemental neutron reflector and as shielding of the lunar outpost from the reactors' neutron and gamma radiation.

A school investigation into Chernobyl fallout

NASA Astrophysics Data System (ADS)

Plant, R. D.

1988-01-01

The nuclear power station operating at Chernobyl, just north of Kiev in the Ukraine, USSR, contains four RBMK reactors operating at 1000 MW each. The RBMK reactor is a graphite moderated light water cooled reactor using low enriched uranium fuel. Early on Saturday 26 April 1986 a serious accident occurred to one of the four reactors resulting in the release of radioactive material, some of which was carried by the wind northwards across Poland and Scandinavia. The Ursuline Convent School at Westgate-on-Sea is situated in a small seaside town on the North Kent coast. On 30 April the background count was measured in the physics laboratory of the school using a Mullard ZP1481 Geiger-Muller tube in conjunction with a Panax scaler.

DIII-D research to address key challenges for ITER and fusion energy

NASA Astrophysics Data System (ADS)

Buttery, R. J.; the DIII-D Team

2015-10-01

DIII-D has made significant advances in the scientific basis for fusion energy. The physics mechanism of resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression is revealed as field penetration at the pedestal top, and reduced coil set operation was demonstrated. Disruption runaway electrons were effectively quenched by shattered pellets; runaway dissipation is explained by pitch angle scattering. Modest thermal quench radiation asymmetries are well described NIMROD modelling. With good pedestal regulation and error field correction, low torque ITER baselines have been demonstrated and shown to be compatible with an ITER test blanket module simulator. However performance and long wavelength turbulence degrade as low rotation and electron heating are approached. The alternative QH mode scenario is shown to be compatible with high Greenwald density fraction, with an edge harmonic oscillation demonstrating good impurity flushing. Discharge optimization guided by the EPED model has discovered a new super H-mode with doubled pedestal height. Lithium injection also led to wider, higher pedestals. On the path to steady state, 1 MA has been sustained fully noninductively with βN = 4 and RMP ELM suppression, while a peaked current profile scenario provides attractive options for ITER and a βN = 5 future reactor. Energetic particle transport is found to exhibit a critical gradient behaviour. Scenarios are shown to be compatible with radiative and snowflake divertor techniques. Physics studies reveal that the transition to H mode is locked in by a rise in ion diamagnetic flows. Intrinsic rotation in the plasma edge is demonstrated to arise from kinetic losses. New 3D magnetic sensors validate linear ideal MHD, but identify issues in nonlinear simulations. Detachment, characterized in 2D with sub-eV resolution, reveals a radiation shortfall in simulations. Future facility development targets burning plasma physics with torque free electron heating, the path to steady state with increased off axis currents, and a new divertor solution for fusion reactors.

DIII-D research to address key challenges for ITER and fusion energy

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

Buttery, Richard J.

DIII-D has made significant advances in the scientific basis for fusion energy. The physics mechanism of resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression is revealed as field penetration at the pedestal top, and reduced coil set operation was demonstrated. Disruption runaway electrons were effectively quenched by shattered pellets; runaway dissipation is explained by pitch angle scattering. Modest thermal quench radiation asymmetries are well described NIMROD modeling. With good pedestal regulation and error field correction, low torque ITER baselines have been demonstrated and shown to be compatible with an ITER test blanket module simulator. However performance and long wavelengthmore » turbulence degrade as low rotation and electron heating are approached. The alternative QH mode scenario is shown to be compatible with high Greenwald density fraction, with an edge harmonic oscillation demonstrating good impurity flushing. Discharge optimization guided by the EPED model has discovered a new super H-mode with doubled pedestal height. Lithium injection also led to wider, higher pedestals. On the path to steady state, 1 MA has been sustained fully non inductively with β N = 4 and RMP ELM suppression, while a peaked current profile scenario provides attractive options for ITER and a β N = 5 future reactor. Energetic particle transport is found to exhibit a critical gradient behavior. Scenarios are shown to be compatible with radiative and snowflake diverter techniques. Physics studies reveal that the transition to H mode is locked in by a rise in ion diamagnetic flows. Intrinsic rotation in the plasma edge is demonstrated to arise from kinetic losses. New 3D magnetic sensors validate linear ideal MHD, but identify issues in nonlinear simulations. Detachment, characterized in 2D with sub-eV resolution, reveals a radiation shortfall in simulations. As a result, future facility development targets burning plasma physics with torque free electron heating, the path to steady state with increased off axis currents, and a new divertor solution for fusion reactors.« less

DIII-D research to address key challenges for ITER and fusion energy

DOE PAGES

Buttery, Richard J.

2015-07-29

DIII-D has made significant advances in the scientific basis for fusion energy. The physics mechanism of resonant magnetic perturbation (RMP) edge localized mode (ELM) suppression is revealed as field penetration at the pedestal top, and reduced coil set operation was demonstrated. Disruption runaway electrons were effectively quenched by shattered pellets; runaway dissipation is explained by pitch angle scattering. Modest thermal quench radiation asymmetries are well described NIMROD modeling. With good pedestal regulation and error field correction, low torque ITER baselines have been demonstrated and shown to be compatible with an ITER test blanket module simulator. However performance and long wavelengthmore » turbulence degrade as low rotation and electron heating are approached. The alternative QH mode scenario is shown to be compatible with high Greenwald density fraction, with an edge harmonic oscillation demonstrating good impurity flushing. Discharge optimization guided by the EPED model has discovered a new super H-mode with doubled pedestal height. Lithium injection also led to wider, higher pedestals. On the path to steady state, 1 MA has been sustained fully non inductively with β N = 4 and RMP ELM suppression, while a peaked current profile scenario provides attractive options for ITER and a β N = 5 future reactor. Energetic particle transport is found to exhibit a critical gradient behavior. Scenarios are shown to be compatible with radiative and snowflake diverter techniques. Physics studies reveal that the transition to H mode is locked in by a rise in ion diamagnetic flows. Intrinsic rotation in the plasma edge is demonstrated to arise from kinetic losses. New 3D magnetic sensors validate linear ideal MHD, but identify issues in nonlinear simulations. Detachment, characterized in 2D with sub-eV resolution, reveals a radiation shortfall in simulations. As a result, future facility development targets burning plasma physics with torque free electron heating, the path to steady state with increased off axis currents, and a new divertor solution for fusion reactors.« less

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Saldikov, I. S.; Ternovykh, M. Yu; Fomichenko, P. A.; Gerasimov, A. S.

2017-01-01

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

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

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

A multi-physics analysis for the actuation of the SSS in opal reactor

NASA Astrophysics Data System (ADS)

Ferraro, Diego; Alberto, Patricio; Villarino, Eduardo; Doval, Alicia

2018-05-01

OPAL is a 20 MWth multi-purpose open-pool type Research Reactor located at Lucas Heights, Australia. It was designed, built and commissioned by INVAP between 2000 and 2006 and it has been operated by the Australia Nuclear Science and Technology Organization (ANSTO) showing a very good overall performance. On November 2016, OPAL reached 10 years of continuous operation, becoming one of the most reliable and available in its kind worldwide, with an unbeaten record of being fully operational 307 days a year. One of the enhanced safety features present in this state-of-art reactor is the availability of an independent, diverse and redundant Second Shutdown System (SSS), which consists in the drainage of the heavy water reflector contained in the Reflector Vessel. As far as high quality experimental data is available from reactor commissioning and operation stages and even from early component design validation stages, several models both regarding neutronic and thermo-hydraulic approaches have been developed during recent years using advanced calculations tools and the novel capabilities to couple them. These advanced models were developed in order to assess the capability of such codes to simulate and predict complex behaviours and develop highly detail analysis. In this framework, INVAP developed a three-dimensional CFD model that represents the detailed hydraulic behaviour of the Second Shutdown System for an actuation scenario, where the heavy water drainage 3D temporal profiles inside the Reflector Vessel can be obtained. This model was validated, comparing the computational results with experimental measurements performed in a real-size physical model built by INVAP during early OPAL design engineering stages. Furthermore, detailed 3D Serpent Monte Carlo models are also available, which have been already validated with experimental data from reactor commissioning and operating cycles. In the present work the neutronic and thermohydraulic models, available for OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

Joint tests at INL and CEA of a transient hot wire needle probe for in-pile thermal conductivity measurement

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

Daw, J.E.; Knudson, D.L.; Villard, J.F.

2015-07-01

Thermal conductivity is a key property that must be known for proper design, testing, and deployment of new fuels and structural materials in nuclear reactors. Thermal conductivity is highly dependent on the physical structure, chemical composition, and the state of the material. Typically, thermal conductivity changes that occur during irradiation are currently measured out-of-pile using a 'cook and look' approach. But repeatedly removing samples from a test reactor to make measurements is expensive, has the potential to disturb phenomena of interest, and only provides understanding of the sample's end state when each measurement is made. There are also limited thermo-physicalmore » property data available for advanced fuels; and such data are needed for simulation codes, the development of next generation reactors, and advanced fuels for existing nuclear plants. Being able to quickly characterize fuel thermal conductivity during irradiation can improve the fidelity of data, reduce costs of post-irradiation examinations, increase understanding of how fuels behave under irradiation, and confirm or improve existing thermal conductivity measurement techniques. This paper discusses efforts to develop and evaluate an innovative in-pile thermal conductivity sensor based on the transient hot wire thermal conductivity method (THWM), using a single needle probe (NP) containing a line heat source and thermocouple embedded in the fuel. The sensor that has been designed and manufactured by the Idaho National Laboratory (INL) includes a unique combination of materials, geometry, and fabrication techniques that make the hot wire method suitable for in-pile applications. In particular, efforts were made to minimize the influence of the sensor and maximize fuel hot-wire heating. The probe has a thermocouple-like construction with high temperature resistant materials that remain ductile while resisting transmutation and materials interactions. THWM-NP prototypes were fabricated for both room temperature proof-of-concept evaluations and high temperature testing. Evaluations have been performed jointly by the INL and the French Alternative Energies and Atomic Energy Commission (CEA), both in Idaho Falls (USA) and in Cadarache (France), in the framework of a collaborative program for instrumentation of Material Testing Reactors. Initial tests were conducted on samples with a large range of thermal conductivities and temperatures ranging from 20 deg. C to 600 deg. C. Particularly, tests were recently performed on a sample having thermal conductivity and dimensions similar to UO{sub 2} and MOX nuclear fuels, in order to validate the ability of this sensor to operate for in-pile characterization of Light Water Reactors fuels. The results of the tests already completed at INL and CEA indicate that the Transient Hot Wire Needle Probe offers an enhanced method for in-pile detection of thermal conductivity. (authors)« less

Energy from nuclear fission()

NASA Astrophysics Data System (ADS)

Ripani, M.

2015-08-01

The main features of nuclear fission as physical phenomenon will be revisited, emphasizing its peculiarities with respect to other nuclear reactions. Some basic concepts underlying the operation of nuclear reactors and the main types of reactors will be illustrated, including fast reactors, showing the most important differences among them. The nuclear cycle and radioactive-nuclear-waste production will be also discussed, along with the perspectives offered by next generation nuclear assemblies being proposed. The current situation of nuclear power in the world, its role in reducing carbon emission and the available resources will be briefly illustrated.

Operating manual for the Bulk Shielding Reactor

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

Not Available

1983-04-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

Operating manual for the Bulk Shielding Reactor

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

Not Available

1987-03-01

The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

Core plasma design of the compact helical reactor with a consideration of the equipartition effect

NASA Astrophysics Data System (ADS)

Goto, T.; Miyazawa, J.; Yanagi, N.; Tamura, H.; Tanaka, T.; Sakamoto, R.; Suzuki, C.; Seki, R.; Satake, S.; Nunami, M.; Yokoyama, M.; Sagara, A.; the FFHR Design Group

2018-07-01

Integrated physics analysis of plasma operation scenario of the compact helical reactor FFHR-c1 has been conducted. The DPE method, which predicts radial profiles in a reactor by direct extrapolation from the reference experimental data, has been extended to implement the equipartition effect. Close investigation of the plasma operation regime has been conducted and a candidate plasma operation point of FFHR-c1 has been identified within the parameter regime that has already been confirmed in LHD experiment in view of MHD equilibrium, MHD stability and neoclassical transport.

Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel

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

Conant, Andrew; Erickson, Anna; Robel, Martin

Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurizedmore » water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240Pu/ 239Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240Pu/ 239Pu and 137Cs/ 135Cs ratios, respectively. Lastly, this study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.« less

Sensitivity and Uncertainty Analysis of Plutonium and Cesium Isotopes in Modeling of BR3 Reactor Spent Fuel

DOE PAGES

Conant, Andrew; Erickson, Anna; Robel, Martin; ...

2017-02-03

Nuclear forensics has a broad task to characterize recovered nuclear or radiological material and interpret the results of investigation. One approach to isotopic characterization of nuclear material obtained from a reactor is to chemically separate and perform isotopic measurements on the sample and verify the results with modeling of the sample history, for example, operation of a nuclear reactor. The major actinide plutonium and fission product cesium are commonly measured signatures of the fuel history in a reactor core. This study investigates the uncertainty of the plutonium and cesium isotope ratios of a fuel rod discharged from a research pressurizedmore » water reactor when the location of the sample is not known a priori. A sensitivity analysis showed overpredicted values for the 240Pu/ 239Pu ratio toward the axial center of the rod and revealed a lower probability of the rod of interest (ROI) being on the periphery of the assembly. The uncertainty analysis found the relative errors due to only the rod position and boron concentration to be 17% to 36% and 7% to 15% for the 240Pu/ 239Pu and 137Cs/ 135Cs ratios, respectively. Lastly, this study provides a method for uncertainty quantification of isotope concentrations due to the location of the ROI. Similar analyses can be performed to verify future chemical and isotopic analyses.« less

Health physics aspects of advanced reactor licensing reviews

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

Hinson, C.S.

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovativemore » design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.« less

PARTIAL ECONOMIC STUDY OF STEAM COOLED HEAVY WATER MODERATED REACTORS

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

None

1960-04-01

Steam-cooled reactors are compared with CAHDU for costs of Calandria tubes, pressure tubes. heavy water moderator, heavy water reflector, fuel supply, heat exchanger, and turbine generator. A direct-cycle lightsteam-cooled heavy- water-moderated pressure-tube reactor formed the basic reactor design for the study. Two methods of steam circulation through the reactor were examined. In both cases the steam was generated outside the reactor and superheated in the reactor core. One method consisted of a series of reactor and steam generator passes. The second method consisted of the Loeffler cycle and its modifications. The fuel was assumed to be natural cylindrical UO/sub 2/more » pellets sheathed in a hypothetical material with the nuclear properties of Zircaloy, but able to function at temperatures to 900 deg F. For the conditions assumed, the longer the rod, the higher the outlet temperature and therefore the higher the efficiency. The turbine cycle efficiency was calculated on the assumption that suitable steam generators are available. As the neutron losses to the pressure tubes were significant, an economic analysis of insulated pressure tubes is included. A description of the physics program for steam-cooled reactors is included. Results indicated that power from the steam-cooled reactor would cost 1.4 mills/ kwh compared with 1.25 mills/kwh for CANDU. (M.C.G.)« less

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

Virginia Finley

The results of the 1999 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and non-radioactive pollutants (if any) that are added to the environment as a result of PPPL's operations. The report also summarizes environmental initiatives, assessments, and programs that were undertaken in 1999. The Princeton Plasma Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program ismore » to create innovations to make fusion power a practical reality--an alternative energy source. 1999 marked the first year of National Spherical Torus Experiment (NSTX) operations and Tokamak Fusion Test Reactor (TFTR) dismantlement and deconstruction activities. A collaboration among fourteen national laboratories, universities, and research institutions, the NSTX is a major element in the U.S. Fusion Energy Sciences Program. It has been designed to test the physics principles of spherical torus (ST) plasmas. The ST concept could play an important role in the development of smaller, more economical fusion reactors. With its completion within budget and ahead of its target schedule, NSTX first plasma occurred on February 12, 1999. The 1999 performance of the Princeton Plasma Physics Laboratory was rated ''outstanding'' by the U.S. Department of Energy in the Laboratory Appraisal report issued early in 2000. The report cited the Laboratory's consistently excellent scientific and technological achievements, its successful management practices, and included high marks in a host of other areas including environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary agreement with the New Jersey Department of Environmental Protection. PPPL monitored for the presence of non-radiological contaminants, mainly volatile organic compounds (components of degreasing solvents). Monitoring revealed the presence of low levels of volatile organic compounds in an area adjacent to PPPL. Also, PPPL's radiological monitoring program characterized the ambient, background levels of tritium in the environment and from the TFTR stack; the data are presented in this report.« less

Characterization of fast neutron spectrum in the TRIGA for hardness testing of electronic components

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

Nelson, George W.

1986-07-01

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

Ceramic Coatings for Clad (The C 3 Project): Advanced Accident-Tolerant Ceramic Coatings for Zr-Alloy Cladding

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

Sickafus, Kurt E.; Wirth, Brian; Miller, Larry

The goal of this NEUP-IRP project is to develop a fuel concept based on an advanced ceramic coating for Zr-alloy cladding. The coated cladding must exhibit demonstrably improved performance compared to conventional Zr-alloy clad in the following respects: During normal service, the ceramic coating should decrease cladding oxidation and hydrogen pickup (the latter leads to hydriding and embrittlement). During a reactor transient (e.g., a loss of coolant accident), the ceramic coating must minimize or at least significantly delay oxidation of the Zr-alloy cladding, thus reducing the amount of hydrogen generated and the oxygen ingress into the cladding. The specific objectivesmore » of this project are as follows: To produce durable ceramic coatings on Zr-alloy clad using two possible routes: (i) MAX phase ceramic coatings or similar nitride or carbide coatings; and (ii) graded interface architecture (multilayer) ceramic coatings, using, for instance, an oxide such as yttria-stabilized zirconia (YSZ) as the outer protective layer. To characterize the structural and physical properties of the coated clad samples produced in 1. above, especially the corrosion properties under simulated normal and transient reactor operating conditions. To perform computational analyses to assess the effects of such coatings on fuel performance and reactor neutronics, and to perform fuel cycle analyses to assess the economic viability of modifying conventional Zr-alloy cladding with ceramic coatings. This project meets a number of the goals outlined in the NEUP-IRP call for proposals, including: Improve the fuel/cladding system through innovative designs (e.g. coatings/liners for zirconium-based cladding) Reduce or eliminate hydrogen generation Increase resistance to bulk steam oxidation Achievement of our goals and objectives, as defined above, will lead to safer light-water reactor (LWR) nuclear fuel assemblies, due to improved cladding properties and built-in accident resistance, as well as the possibilities for enhanced fuel/clad system performance and longevity.« less

Mars Atmospheric Conversion to Methane and Water: An Engineering Model of the Sabatier Reactor with Characterization of Ru/Al2O3 for Long Duration Use on Mars

NASA Technical Reports Server (NTRS)

Meier, Anne J.; Shah, Malay; Petersen, Elspeth; Hintze, Paul; Muscatello, Tony

2017-01-01

The Atmospheric Processing Module (APM) is a Mars In-Situ Resource Utilization (ISRU) technology designed to demonstrate conversion of the Martian atmosphere into methane and water. The Martian atmosphere consists of approximately 95 carbon dioxide (CO2) and residual argon and nitrogen. APM utilizes cryocoolers for CO2 acquisition from a simulated Martian atmosphere and pressure. The captured CO2 is sublimated and pressurized as a feedstock into the Sabatier reactor, which converts CO2 and hydrogen to methane and water. The Sabatier reaction occurs over a packed bed reactor filled with Ru/Al2O3 pellets. The long duration use of the APM system and catalyst was investigated for future scaling and failure limits. Failure of the catalyst was detected by gas chromatography and temperature sensors on the system. Following this, characterization and experimentation with the catalyst was carried out with analysis including x-ray photoelectron spectroscopy and scanning electron microscopy with elemental dispersive spectroscopy. This paper will discuss results of the catalyst performance, the overall APM Sabatier approach, as well as intrinsic catalyst considerations of the Sabatier reactor performance incorporated into a chemical model.

Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

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

Daniel, W. E.; Rudisill, T. S.; O'Rourke, P. E.

2017-01-26

In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgasmore » composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.« less

Acetic acid effects on enhancement of growth rate and reduction of amorphous carbon deposition on CNT arrays along a growth window in a floating catalyst reactor

NASA Astrophysics Data System (ADS)

Maghrebi, Morteza; Khodadadi, Abbas Ali; Mortazavi, Yadollah; Sane, Ali; Rahimi, Mohsen; Shirazi, Yaser; Tsakadze, Zviad; Mhaisalkar, Subodh

2009-11-01

The mm-long carbon nanotube (CNT) arrays were grown in a floating catalyst reactor, using xylene-ferrocene and a small amount of acetic acid as the feed. The CNT arrays deposited on a quartz substrate at several positions along the reactor were extensively characterized using Raman spectroscopy, scanning electron microscopy, X-ray diffraction, high-resolution transmission electron microscopy, and optical microscopy. Various characterization methods consistently reveal that the acetic acid additive to the feed alleviates deposition of amorphous carbon layer, which gradually thickens CNTs along the reactor. The acetic acid also resulted in a higher growth rate along the so-called growth window, where CNT arrays are deposited on the quartz substrate. High-performance liquid chromatography of extracted byproducts (PAHs) confirmed the presence of some polycyclic aromatic hydrocarbons. The solid weight of PAHs decreased upon addition of ferrocene as the catalyst precursor, as well as of acetic acid to xylene feed. The results suggest that primary light products of xylene pyrolysis can be competitive reactants for both catalytic and subsequent pyrolytic reactions. They may also be more efficient feeds for CNT growth than xylene itself.

Nonlinear ultrasonic characterization of precipitation in 17-4PH stainless steel

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

Matlack, Kathryn; Bradley, Harrison A.; Thiele, Sebastian

2015-04-01

The extension of operational lifetime of most US nuclear reactors will cause reactor pressure vessel to be exposed to increased levels of neutron radiation damage. This research is part of a broader effort to develop a nondestructive evaluation technique to monitor radiation damage in reactor pressure vessel steels. The main contributor to radiation embrittlement in these steels is the formation of copper-rich precipitates. In this work, a precipitate hardenable martensitic alloy, 17-4PH stainless steel is exposed to thermal aging treatments, and used as a surrogate material to study the effects of copper precipitates on the measured acoustic nonlinearity parameter. Previousmore » work has demonstrated the effectiveness of these nonlinear ultrasonic (NLU) measurements in the characterization of radiation-induced microstructural changes in neutron irradiated reactor pressure vessel steels. NLU measurements using Rayleigh surface waves are performed on 17-4PH samples subjected to isothermal aging. NLU measurements are interpreted with hardness, thermo-electric power, TEM, and atom probe tomography measurements. The Rayleigh wave measurements showed a decrease in the acoustic nonlinearity parameter with increasing aging time, consistent with evidence of increasing number density of nucleated precipitates.« less

Characteristic of molten fluoride salt system LiF-BeF2 (Flibe) and LiF-NaF-KF (Flinak) as coolant and fuel carrier in molten salt reactor (MSR)

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Al-Areqi, Wadee'ah Mohd; Ruf, Mohd'Izzat Fahmi Mohd; Majid, Amran Ab.

2017-01-01

Interest of fluoride salts have recently revived due to the high temperature application in nuclear reactors. Molten Salt Reactor (MSR) was designed to operate at high temperature in range 700 - 800°C and its fuel is dissolved in a circulating molten fluoride salt mixture. Molten fluoride salts are stable at high temperature, have good heat transfer properties and can dissolve high concentration of actinides and fission product. The aim of this paper was to discuss the physical properties (melting temperature, density and heat capacity) of two systems fluoride salt mixtures i.e; LiF-BeF2 (Flibe) and LiF-NaF-KF (Flinak) in terms of their application as coolant and fuel solvent in MSR. Both of these salts showed almost same physical properties but different applications in MSR. The advantages and the disadvantages of these fluoride salt systems will be discussed in this paper.

SlimCS—compact low aspect ratio DEMO reactor with reduced-size central solenoid

NASA Astrophysics Data System (ADS)

Tobita, K.; Nishio, S.; Sato, M.; Sakurai, S.; Hayashi, T.; Shibama, Y. K.; Isono, T.; Enoeda, M.; Nakamura, H.; Sato, S.; Ezato, K.; Hayashi, T.; Hirose, T.; Ide, S.; Inoue, T.; Kamada, Y.; Kawamura, Y.; Kawashima, H.; Koizumi, N.; Kurita, G.; Nakamura, Y.; Mouri, K.; Nishitani, T.; Ohmori, J.; Oyama, N.; Sakamoto, K.; Suzuki, S.; Suzuki, T.; Tanigawa, H.; Tsuchiya, K.; Tsuru, D.

2007-08-01

The concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (βN), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high nGW (because of an increase in Ip), which allows efficient use of the capacity of high βN. From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

Delayed neutron spectral data for Hansen-Roach energy group structure

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

Campbell, J.M.; Spriggs, G.D.

A detailed knowledge of delayed neutron spectra is important in reactor physics. It not only allows for an accurate estimate of the effective delayed neutron fraction {beta}{sub eff} but also is essential to calculating important reactor kinetic parameters, such as effective group abundances and the ratio of {beta}{sub eff} to the prompt neutron generation time. Numerous measurements of delayed neutron spectra for various delayed neutron precursors have been performed and reported in the literature. However, for application in reactor physics calculations, these spectra are usually lumped into one of the traditional six groups of delayed neutrons in accordance to theirmore » half-lives. Subsequently, these six-group spectra are binned into energy intervals corresponding to the energy intervals of a chosen nuclear cross-section set. In this work, the authors present a set of delayed neutron spectra that were formulated specifically to match Keepin`s six-group parameters and the 16-energy-group Hansen-Roach cross sections.« less

Filtered epithermal quasi-monoenergetic neutron beams at research reactor facilities.

PubMed

Mansy, M S; Bashter, I I; El-Mesiry, M S; Habib, N; Adib, M

2015-03-01

Filtered neutron techniques were applied to produce quasi-monoenergetic neutron beams in the energy range of 1.5-133keV at research reactors. A simulation study was performed to characterize the filter components and transmitted beam lines. The filtered beams were characterized in terms of the optimal thickness of the main and additive components. The filtered neutron beams had high purity and intensity, with low contamination from the accompanying thermal emission, fast neutrons and γ-rays. A computer code named "QMNB" was developed in the "MATLAB" programming language to perform the required calculations. Copyright © 2014 Elsevier Ltd. All rights reserved.

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

Finley, V.L. and Levine, J.D.

The results of the 1997 environmental surveillance and monitoring program for the Princeton Plasma Physics Laboratory (PPPL) are presented and discussed. The purpose of this report is to provide the U.S. Department of Energy and the public with information on the level of radioactive and non-radioactive pollutants, if any, that are added to the environment as a result of PPPL's operations. During Calendar Year 1997, PPPL's Tokamak Fusion Test Reactor (TFTR) completed fifteen years of fusion experiments begun in 1982. Over the course of three and half years of deuterium-tritium (D-T) plasma experiments, PPPL set a world record of 10.7more » million watts of controlled fusion power, more than 700 tritium shots pulsed into the reactor vessel generating more than 5.6 x 10 20 neutron and 1.6 gigajoules of fusion energy and researchers studied plasma science experimental data, which included "enhanced reverse shear techniques." As TFTR was completing its historic operations, PPPL participated with the Oak Ridge National Laboratory, Columbia University, and the University of Washington (Seattle) in a collaboration effort to design the National Spherical Torus Experiment (NSTX). This next device, NSTX, is located in the former TFTR Hot Cell on D site, and it is designed to be a smaller and more economical torus fusion reactor. Construction of this device began in late 1997, and first plasma in scheduled for early 1999. For 1997, the U.S. Department of Energy in its Laboratory Appraisal report rated the overall performance of Princeton Plasma Physics Laboratory as "excellent." The report cited the Laboratory's consistently excellent scientific and technological achievements and its successful management practices, which included high marks for environmental management, employee health and safety, human resources administration, science education, and communications. Groundwater investigations continued under a voluntary agreement with the New Jersey Department of Environmental Protection. PPPL monitored the presence of non-radiological contaminants, mainly volatile organic compounds (components of degreasing solvents). Monitoring revealed the presence of low levels of volatile organic compounds in an adjacent area to PPPL. Also, PPPL's radiological monitoring program characterized the ambient, background levels of tritium in the environment and from the TFTR stack; the data are presented in this report.« less

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

NASA Astrophysics Data System (ADS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T. J.; Cesar, J.; Cushman, P.; Dent, J. B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H. R.; Hays, C. C.; Iyer, V.; Jastram, A.; Kadribasic, F.; Kennedy, A.; Kubik, A.; Lang, K.; Mahapatra, R.; Mandic, V.; Marianno, C.; Martin, R. D.; Mast, N.; McDeavitt, S.; Mirabolfathi, N.; Mohanty, B.; Nakajima, K.; Newhouse, J.; Newstead, J. L.; Ogawa, I.; Phan, D.; Proga, M.; Rajput, A.; Roberts, A.; Rogachev, G.; Salazar, R.; Sander, J.; Senapati, K.; Shimada, M.; Soubasis, B.; Strigari, L.; Tamagawa, Y.; Teizer, W.; Vermaak, J. I. C.; Villano, A. N.; Walker, J.; Webb, B.; Wetzel, Z.; Yadavalli, S. A.

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5-20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

Treatment of low strength industrial cluster wastewater by anaerobic hybrid reactor.

PubMed

Kumar, Amit; Yadav, Asheesh Kumar; Sreekrishnan, T R; Satya, Santosh; Kaushik, C P

2008-05-01

The study was aimed at treating the complex, combined wastewater generated in Mangolpuri industrial cluster. It was considered as a low strength wastewater with respect to its organic content. Anaerobic treatment of this wastewater was studied using an anaerobic hybrid reactor (AHR) which combined the best features of both the upflow anaerobic sludge blanket (UASB) reactor and anaerobic fluidized bed rector (AFBR). The performance of the reactor under different organic and hydraulic loading rates were studied. The COD removal reached 94% at an organic loading rate (OLR) of 2.08 kg COD m(-3)d(-1) at an hydraulic retention time (HRT) of 6.0 h. The granules developed were characterized in terms of their diameter and terminal settling velocity.

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

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

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

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

Yung, Matthew M.; Stanton, Alexander R.; Iisa, Kristiina

Metal-impregnated (Ni or Ga) ZSM-5 catalysts were studied for biomass pyrolysis vapor upgrading to produce hydrocarbons using three reactors constituting a 100 000x change in the amount of catalyst used in experiments. Catalysts were screened for pyrolysis vapor phase upgrading activity in two small-scale reactors: (i) a Pyroprobe with a 10 mg catalyst in a fixed bed and (ii) a fixed-bed reactor with 500 mg of catalyst. The best performing catalysts were then validated with a larger scale fluidized-bed reactor (using ~1 kg of catalyst) that produced measurable quantities of bio-oil for analysis and evaluation of mass balances. Despite somemore » inherent differences across the reactor systems (such as residence time, reactor type, analytical techniques, mode of catalyst and biomass feed) there was good agreement of reaction results for production of aromatic hydrocarbons, light gases, and coke deposition. Relative to ZSM-5, Ni or Ga addition to ZSM-5 increased production of fully deoxygenated aromatic hydrocarbons and light gases. In the fluidized bed reactor, Ga/ZSM-5 slightly enhanced carbon efficiency to condensed oil, which includes oxygenates in addition to aromatic hydrocarbons, and reduced oil oxygen content compared to ZSM-5. Ni/ZSM-5, while giving the highest yield of fully deoxygenated aromatic hydrocarbons, gave lower overall carbon efficiency to oil but with the lowest oxygen content. Reaction product analysis coupled with fresh and spent catalyst characterization indicated that the improved performance of Ni/ZSM-5 is related to decreasing deactivation by coking, which keeps the active acid sites accessible for the deoxygenation and aromatization reactions that produce fully deoxygenated aromatic hydrocarbons. The addition of Ga enhances the dehydrogenation activity of the catalyst, which leads to enhanced olefin formation and higher fully deoxygenated aromatic hydrocarbon yields compared to unmodified ZSM-5. Catalyst characterization by ammonia temperature programmed desorption, surface area measurements, and postreaction temperature-programmed oxidation (TPO) also showed that the metal-modified zeolites retained a greater percentage of their initial acidity and surface area, which was consistent between the reactor scales. These results demonstrate that the trends observed with smaller (milligram to gram) catalyst reactors are applicable to larger, more industrially relevant (kg) scales to help guide catalyst research toward application.« less

Characterization of elemental release during microbe granite interactions at T = 28 °C

NASA Astrophysics Data System (ADS)

Wu, Lingling; Jacobson, Andrew D.; Hausner, Martina

2008-02-01

This study used batch reactors to characterize the mechanisms and rates of elemental release (Al, Ca, K, Mg, Na, F, Fe, P, Sr, and Si) during interaction of a single bacterial species ( Burkholderia fungorum) with granite at T = 28 °C for 35 days. The objective was to evaluate how actively metabolizing heterotrophic bacteria might influence granite weathering on the continents. We supplied glucose as a C source, either NH 4 or NO 3 as N sources, and either dissolved PO 4 or trace apatite in granite as P sources. Cell growth occurred under all experimental conditions. However, solution pH decreased from ˜7 to 4 in NH 4-bearing reactors, whereas pH remained near-neutral in NO 3-bearing reactors. Measurements of dissolved CO 2 and gluconate together with mass-balances for cell growth suggest that pH lowering in NH 4-bearing reactors resulted from gluconic acid release and H + extrusion during NH 4 uptake. In NO 3-bearing reactors, B. fungormum likely produced gluconic acid and consumed H + simultaneously during NO 3 utilization. Over the entire 35-day period, NH 4-bearing biotic reactors yielded the highest release rates for all elements considered. However, chemical analyses of biomass show that bacteria scavenged Na, P, and Sr during growth. Abiotic control reactors followed different reaction paths and experienced much lower elemental release rates compared to biotic reactors. Because release rates inversely correlate with pH, we conclude that proton-promoted dissolution was the dominant reaction mechanism. Solute speciation modeling indicates that formation of Al-F and Fe-F complexes in biotic reactors may have enhanced mineral solubilities and release rates by lowering Al and Fe activities. Mass-balances further reveal that Ca-bearing trace phases (calcite, fluorite, and fluorapatite) provided most of the dissolved Ca, whereas more abundant phases (plagioclase) contributed negligible amounts. Our findings imply that during the incipient stages of granite weathering, heterotrophic bacteria utilizing glucose and NH 4 only moderately elevate silicate weathering reactions that consume atmospheric CO 2. However, by enhancing the dissolution of non-silicate, Ca-bearing trace minerals, they could contribute to high Ca/Na ratios commonly observed in granitic watersheds.

Co-gasification of coal and biomass: Synergy, characterization and reactivity of the residual char.

PubMed

Hu, Junhao; Shao, Jingai; Yang, Haiping; Lin, Guiying; Chen, Yingquan; Wang, Xianhua; Zhang, Wennan; Chen, Hanping

2017-11-01

The synergy effect between coal and biomass in their co-gasification was studied in a vertical fixed bed reactor, and the physic-chemical structural characteristics and gasification reactivity of the residual char obtained from co-gasification were also investigated. The results shows that, conversion of the residual char and tar into gas is enhanced due to the synergy effect between coal and biomass. The physical structure of residual char shows more pore on coal char when more biomass is added in the co-gasification. The migration of inorganic elements between coal and biomass was found, the formation and competitive role of K 2 SiO 3 , KAlSiO 4 , and Ca 3 Al 2 (SiO 4 ) 3 is a mechanism behind the synergy. The graphization degree is enhanced but size of graphite crystallite in the residual char decreases with biomass blending ratio increasing. TGA results strongly suggest the big difference in the reactivity of chars derived from coal and biomass in spite of influence from co-gasification. Copyright © 2017 Elsevier Ltd. All rights reserved.

Highly attrition-resistant zinc oxide-based sorbents for H{sub 2}S removal by spray-drying technique

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

Lee, J.B.; Baek, J.I.; Ryu, C.K.

2008-07-15

A ZnO-based sorbent, ZAC 32N, applicable to transport reactors was successfully prepared by the spray-drying technique. Another sorbent, ZAC 32SU, was prepared by scale-up preparation of ZAC 32N sorbent. The physical properties of the sorbents such as attrition resistance, specific surface area, pore volume, and particle size were extensively characterized and exhibited a good potential for use in transport applications. The chemical reactivity tested in the thermogravimetric analyzer and microreactor exhibited desirable characteristics for effective desulfurization of syngas streams in the range of 450-550{sup o}C. Bench-scale tests for the sorbent ZAC 32SU were performed for a continuous 160 h withmore » a steady solid circulation of 54.6 kg/h. The results showed 99.5%+ desulfurization at 500-550{sup o}C and reasonable regenerability at 550-620{sup o}C. Test results on the physical properties and chemical reactivity indicated that the performance of developed sorbents proved to be outstanding.« less

Radiation Resistance of the U(Al, Si)3 Alloy: Ion-Induced Disordering

PubMed Central

Yaniv, Gili; Horak, Pavel; Vacik, Jiri; Mykytenko, Natalia; Rafailov, Gennady; Dahan, Itzchak; Fuks, David; Kiv, Arik

2018-01-01

During the exploitation of nuclear reactors, various U-Al based ternary intermetallides are formed in the fuel-cladding interaction layer. Structure and physical properties of these intermetallides determine the radiation resistance of cladding and, ultimately, the reliability and lifetime of the nuclear reactor. In current research, U(Al, Si)3 composition was studied as a potential constituent of an interaction layer. Phase content of the alloy of an interest was ordered U(Al, Si)3, structure of which was reported earlier, and pure Al (constituting less than 20 vol % of the alloy). This alloy was investigated prior and after the irradiation performed by Ar ions at 30 keV. The irradiation was performed on the transmission electron microscopy (TEM, JEOL, Japan) samples, characterized before and after the irradiation process. Irradiation induced disorder accompanied by stress relief. Furthermore, it was found that there is a dose threshold for disordering of the crystalline matter in the irradiated region. Irradiation at doses equal or higher than this threshold resulted in almost solely disordered phase. Using the program “Stopping and Range of Ions in Matter” (SRIM), the parameters of penetration of Ar ions into the irradiated samples were estimated. Based on these estimations, the dose threshold for ion-induced disordering of the studied material was assessed. PMID:29393870

SCALE Code System 6.2.2

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

Rearden, Bradley T.; Jessee, Matthew Anderson

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules including 3 deterministic and 3 Monte Carlomore » radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results. SCALE 6.2 represents one of the most comprehensive revisions in the history of SCALE, providing several new capabilities and significant improvements in many existing features.« less

Radiation Resistance of the U(Al, Si)₃ Alloy: Ion-Induced Disordering.

PubMed

Meshi, Louisa; Yaniv, Gili; Horak, Pavel; Vacik, Jiri; Mykytenko, Natalia; Rafailov, Gennady; Dahan, Itzchak; Fuks, David; Kiv, Arik

2018-02-02

During the exploitation of nuclear reactors, various U-Al based ternary intermetallides are formed in the fuel-cladding interaction layer. Structure and physical properties of these intermetallides determine the radiation resistance of cladding and, ultimately, the reliability and lifetime of the nuclear reactor. In current research, U(Al, Si)₃ composition was studied as a potential constituent of an interaction layer. Phase content of the alloy of an interest was ordered U(Al, Si)₃, structure of which was reported earlier, and pure Al (constituting less than 20 vol % of the alloy). This alloy was investigated prior and after the irradiation performed by Ar ions at 30 keV. The irradiation was performed on the transmission electron microscopy (TEM, JEOL, Japan) samples, characterized before and after the irradiation process. Irradiation induced disorder accompanied by stress relief. Furthermore, it was found that there is a dose threshold for disordering of the crystalline matter in the irradiated region. Irradiation at doses equal or higher than this threshold resulted in almost solely disordered phase. Using the program "Stopping and Range of Ions in Matter" (SRIM), the parameters of penetration of Ar ions into the irradiated samples were estimated. Based on these estimations, the dose threshold for ion-induced disordering of the studied material was assessed.

Characterization of modified zeolite as microbial immobilization media on POME anaerobic digestion

NASA Astrophysics Data System (ADS)

Cahyono, Rochim B.; Ismiyati, Sri; Ginting, Simparmin Br; Mellyanawaty, Melly; Budhijanto, Wiratni

2018-03-01

As the world’s biggest palm oil producer, Indonesia generates also huge amount of Palm Oil Mill Effluent (POME) wastewater and causes serious problem in environment. In conventional method, POME was converted into biogas using lagoon system which required extensive land area. Anaerobic Fluidized Bed Reactor (AFBR) proposes more effective biogas producing with smaller land area. In the proposed system, a immobilization media would be main factor for enhancing productivity. This research studied on characterization of Lampung natural zeolite as immobilization media in the AFBR system for POME treatment. Various activation method such as physical and chemical were attempted to create more suitable material which has larger surface area, pore size distribution as well as excellent surface structures. The physical method was applied by heating up the material till 400°C while HCl was used on the chemical activation. Based on the result, the chemical activation increased the surface area significantly into 71 m2/g compared to physical as well as original zeolite. The strong acid material was quite effective to enforce the impurities within zeolite pore structure compared to heating up the material. According to distribution data, the Lampung zeolite owned the pore size with the range of 3 – 5 μm which was mesopore material. The pore size was appropriate for immobilization media as it was smaller than size of biogas microbial. The XRD patterns verified that chemical activation could maintain the zeolite structure as the original. Obviously, the SEM photograph showed apparent structure and pore size on the modified zeolite using chemical method. The testing of modified zeolite on the batch system was done to evaluate the characterization process. The modified zeolite using chemical process resulted fast reduction of COD and stabilized the volatile fatty acid as the intermediate product of anaerobic digestion, especially in the beginning of the process. Therefore, the chemical activation process was most suitable to produce the immobilization media from Lampung natural zeolite for POME waste treatment

Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

NASA Astrophysics Data System (ADS)

Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.

2017-12-01

The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.

The moving-ring field-reversed mirror prototype reactor

NASA Astrophysics Data System (ADS)

Smith, A. C., Jr.; Carlson, G. A.; Fleischmann, H. H.; Grossman, W., Jr.; Kammash, T.; Schultz, K. R.; Woodall, D. M.

1981-03-01

A prototype fusion reactor was designed based on magnetic field reversed plasma confinement. A set of physics, technology, and mechanical design criteria were developed in order to make this concept attractive. Six major criteria guide the commercial prototype design. The prototype must: (1) produce net electricity decisively P sub net 70% of P sub gross; (2) scale to an economical commercial plant and have small physical size; (3) have all features required of a correcial upgrade plant (H-3 breeding, etc.); (4) minimize exotic technology and maintenance complexity; (5) promise significantly lower safety hazards than fission plants (environmentally and socially acceptable); and (6) be modular in design to permit repetitive production of components.

1986 Nuclear Science Symposium, 33rd, and 1986 Symposium on Nuclear Power Systems, 18th, Washington, DC, Oct. 29-31, 1986, Proceedings

NASA Technical Reports Server (NTRS)

Stubblefield, F. W. (Editor)

1987-01-01

Papers are presented on space, low-energy physics, and general nuclear science instrumentations. Topics discussed include data acquisition systems and circuits, nuclear medicine imaging and tomography, and nuclear radiation detectors. Consideration is given to high-energy physics instrumentation, reactor systems and safeguards, health physics instrumentation, and nuclear power systems.

Lunar Resource Utilization: Development of a Reactor for Volatile Extraction from Regolith

NASA Technical Reports Server (NTRS)

Kleinhenz, Julie E.; Sacksteder, Kurt R.; Nayagam, Vedha

2007-01-01

The extraction and processing of planetary resources into useful products, known as In- Situ Resource Utilization (ISRU), will have a profound impact on the future of planetary exploration. One such effort is the RESOLVE (Regolith and Environment Science, Oxygen and Lunar Volatiles Extraction) Project, which aims to extract and quantify these resources. As part of the first Engineering Breadboard Unit, the Regolith Volatiles Characterization (RVC) reactor was designed and built at the NASA Glenn Research Center. By heating and agitating the lunar regolith, loosely bound volatiles, such as hydrogen and water, are released and stored in the reactor for later analysis and collection. Intended for operation on a robotic rover, the reactor features a lightweight, compact design, easy loading and unloading of the regolith, and uniform heating of the regolith by means of vibrofluidization. The reactor performance was demonstrated using regolith simulant, JSC1, with favorable results.

Microstructural characterization of a thin film ZrN diffusion barrier in an As-fabricated U-7Mo/Al matrix dispersion fuel plate

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Perez, Emmanuel; Wiencek, Tom; Leenaers, Ann; Van den Berghe, Sven

2015-03-01

The United States High Performance Research Reactor Fuel Development program is developing low enriched uranium fuels for application in research and test reactors. One concept utilizes U-7 wt.% Mo (U-7Mo) fuel particles dispersed in Al matrix, where the fuel particles are coated with a 1 μm-thick ZrN coating. The ZrN serves as a diffusion barrier to eliminate a deleterious reaction that can occur between U-7Mo and Al when a dispersion fuel is irradiated under aggressive reactor conditions. To investigate the final microstructure of a physically-vapor-deposited ZrN coating in a dispersion fuel plate after it was fabricated using a rolling process, characterization samples were taken from a fuel plate that was fabricated at 500 °C using ZrN-coated U-7Mo particles, Al matrix and AA6061 cladding. Scanning electron and transmission electron microscopy analysis were performed. Data from these analyses will be used to support future microstructural examinations of irradiated fuel plates, in terms of understanding the effects of irradiation on the ZrN microstructure, and to determine the role of diffusion barrier microstructure in eliminating fuel/matrix interactions during irradiation. The as-fabricated coating was determined to be cubic-ZrN (cF8) phase. It exhibited a columnar microstructure comprised of nanometer-sized grains and a region of relatively high porosity, mainly near the Al matrix. Small impurity-containing phases were observed at the U-7Mo/ZrN interface, and no interaction zone was observed at the ZrN/Al interface. The bonding between the U-7Mo and ZrN appeared to be mechanical in nature. A relatively high level of oxygen was observed in the ZrN coating, extending from the Al matrix in the ZrN coating in decreasing concentration. The above microstructural characteristics are discussed in terms of what may be most optimal for a diffusion barrier in a dispersion fuel plate application.

Dealing with Historical Discrepancies: The Recovery of National Research Experiment (NRX) Reactor Fuel Rods at Chalk River Laboratories (CRL) - 13324

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

Vickerd, Meggan

2013-07-01

Following the 1952 National Research Experiment (NRX) Reactor accident, fuel rods which had short irradiation histories were 'temporarily' buried in wooden boxes at the 'disposal grounds' during the cleanup effort. The Nuclear Legacy Liabilities Program (NLLP), funded by Natural Resources Canada (NRCan), strategically retrieves legacy waste and restores lands affected by Atomic Energy of Canada Limited (AECL) early operations. Thus under this program the recovery of still buried NRX reactor fuel rods and their relocation to modern fuel storage was identified as a priority. A suspect inventory of NRX fuels was compiled from historical records and various research activities. Sitemore » characterization in 2005 verified the physical location of the fuel rods and determined the wooden boxes they were buried in had degraded such that the fuel rods were in direct contact with the soil. The fuel rods were recovered and transferred to a modern fuel storage facility in 2007. Recovered identification tags and measured radiation fields were used to identify the inventory of these fuels. During the retrieval activity, a discrepancy was discovered between the anticipated number of fuel rods and the number found during the retrieval. A total of 32 fuel rods and cans of cut end pieces were recovered from the specified site, which was greater than the anticipated 19 fuel rods and cans. This discovery delayed the completion of the project, increased the associated costs, and required more than anticipated storage space in the modern fuel storage facility. A number of lessons learned were identified following completion of this project, the most significant of which was the potential for discrepancies within the historical records. Historical discrepancies are more likely to be resolved by comprehensive historical record searches and site characterizations. It was also recommended that a complete review of the wastes generated, and the total affected lands as a result of this historic 1952 NRX accident be undertaken. These lessons and recommendations have lead to changes in how the NLLP is executed in the CRL waste management areas. (authors)« less

Preparation macroconstants to simulate the core of VVER-1000 reactor

NASA Astrophysics Data System (ADS)

Seleznev, V. Y.

2017-01-01

Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.

Rotating Fluidized Bed Reactor for Space Nuclear Propulsion. Annual Report; Design Studies and Experimental Results

NASA Technical Reports Server (NTRS)

1971-01-01

The rotating fluidized bed reactor concept is being investigated for possible application in nuclear propulsion systems. Physics calculations show U-233 to be superior to U-235 as a fuel for a cavity reactor of this type. Preliminary estimates of the effect of hydrogen in the reactor, reflector material, and power peaking are given. A preliminary engineering analysis was made for U-235 and U-233 fueled systems. An evaluation of the parameters affecting the design of the system is given, along with the thrust-to-weight ratios. The experimental equipment is described, as are the special photographic techniques and procedures. Characteristics of the fluidized bed and experimental results are given, including photographic evidence of bed fluidization at high rotational velocities.

Flow reversal power limit for the HFBR

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

Cheng, Lap Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Uncertainties about the afterheat removal capability during the flow reversal has limited the reactor operating power to 30 MW. An experimental and analytical program to address these uncertainties is described in this report. The experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safemore » operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW.« less

Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

NASA Astrophysics Data System (ADS)

Blanchet, David; Fontaine, Bruno

2017-09-01

The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

EBT reactor systems analysis and cost code: description and users guide (Version 1)

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

Santoro, R.T.; Uckan, N.A.; Barnes, J.M.

1984-06-01

An ELMO Bumpy Torus (EBT) reactor systems analysis and cost code that incorporates the most recent advances in EBT physics has been written. The code determines a set of reactors that fall within an allowed operating window determined from the coupling of ring and core plasma properties and the self-consistent treatment of the coupled ring-core stability and power balance requirements. The essential elements of the systems analysis and cost code are described, along with the calculational sequences leading to the specification of the reactor options and their associated costs. The input parameters, the constraints imposed upon them, and the operatingmore » range over which the code provides valid results are discussed. A sample problem and the interpretation of the results are also presented.« less

Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing 235U, 233U, and 232Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

NASA Astrophysics Data System (ADS)

Ioffe, B. L.; Kochurov, B. P.

2012-02-01

A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of 235U. It operates in the open-cycle mode involving 233U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

AGC-2 Irradiation Report

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

Rohrbaugh, David Thomas; Windes, William; Swank, W. David

The Next Generation Nuclear Plant (NGNP) will be a helium-cooled, very high temperature reactor (VHTR) with a large graphite core. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs.[ , ] Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphites have been developed and are considered suitable candidates for the new NGNP reactor design. To support the design and licensing of NGNP core components within a commercial reactor, a completemore » 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 related to the life limiting effects of irradiation creep on key physical properties of the NGNP candidate graphites. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change. In this case, the graphite physically changes i.e., first shrinking and then expanding with increasing neutron dose. This disparity in material volume change can induce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphites is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment.« less

Year One Summary of X-energy Pebble Fuel Development at ORNL

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

Helmreich, Grant W.; Hunn, John D.; McMurray, Jake W.

2017-06-01

The Advanced Reactor Concepts X-energy (ARC-Xe) Pebble Fuel Development project at Oak Ridge National Laboratory (ORNL) has successfully completed its first year, having made excellent progress in accomplishing programmatic objectives. The primary focus of research at ORNL in support of X-energy has been the training of X-energy fuel fabrication engineers and the establishment of US pebble fuel production capabilities able to supply the Xe-100 pebble-bed reactor. These efforts have been strongly supported by particle fuel fabrication and characterization expertise present at ORNL from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program.

Process intensification for the production of hydroxyapatite nanoparticles

NASA Astrophysics Data System (ADS)

Castro, Filipa Juliana Fernandes

Precipitation processes are widely used in chemical industry for the production of particulate solids. In these processes, the chemical and physical nature of synthesized particles is of key importance. The traditional stirred tank batch reactors are affected by non-uniform mixing of reactants, often resulting in broad particle size distribution. The main objective of this thesis was to apply meso and microreactors for the synthesis of hydroxyapatite (HAp) nanoparticles under near-physiological conditions of pH and temperature, in order to overcome the limitations associated with stirred tank batch reactors. Meso and microreactors offer unique features in comparison with conventional chemical reactors. Their high surface-to-volume ratio enables enhanced heat and mass transfer, as well as rapid and efficient mixing. In addition to low consumption of reagents, meso and microreactors are usually operated in continuous flow, making them attractive tools for high throughput experimentation. Precipitation of HAp was first studied in a stirred tank batch reactor, mixing being assured by a novel metal stirrer. HAp was synthetized by mixing diluted aqueous solutions of calcium hydroxide and orthophosphoric acid at 37 °C. After process optimization, a suspension of HAp nanoparticles with pH close to 7 was obtained for a mixing molar ratio Ca/P=1.33. The precipitation process was characterized by three stages: precipitation of amorphous calcium phosphate, transformation of amorphous calcium phosphate into HAp and growth of HAp crystals. The reaction system was further characterized based on equilibrium equations. The resolution of the system, which was possible with the knowledge of three process variables (temperature, pH and calcium concentration), allowed identifying and quantifying all the chemical species present in solution. The proposed model was validated by comparing the experimental and theoretical conductivity. Precipitation of HAp was then investigated in a meso oscillatory flow reactor (meso-OFR). The mesoreactor was first operated batchwise in a vertical tube and experiments were performed under the same conditions of temperature, reactants concentration and power density applied in the stirred tank batch reactor. Despite hydrodynamic conditions being not directly comparable, it was possible to assess the effectiveness of both reactors in terms of mixing and quality of the precipitated particles. The experimental results show the advantages of the meso-OFR over the stirred tank due to the production, about four times faster, of smaller and more uniform HAp nanoparticles. Afterwards, continuous-flow precipitation of HAp was carried out in one meso-OFR and in a series of eight meso-OFRs. Experiments were carried out using fixed frequency (f) and amplitude (x0), varying only the residence time. HAp nanoparticles were successfully obtained in both systems, mean particle size and aggregation degree of the prepared HAp particles decreasing with decreasing residence time. In the present work continuous-flow precipitation of HAp was also investigated in two ultrasonic microreactors. Initially, the process was carried out in a tubular microreactor immersed in an ultrasonic bath, where single-phase (laminar) and gas-liquid flow experiments were both performed. Continuous-flow precipitation of HAp in single-phase flow was then done in a Teflon microreactor with integrated piezoelectric actuator. Rod-like shape HAp nanoparticles were yielded in both reactors under near-physiological conditions of pH and temperature. Further, particles showed improved characteristics, namely in terms of size, shape, particle aggregation and crystallinity. In summary, scale-down of the HAp precipitation process has resulted in the formation of HAp nanoparticles with improved characteristics when compared with HAp particles prepared in a stirred tank batch reactor. Therefore, we believe that the work developed can be a useful contribution to the development of a platform for the continuous production of high quality HAp nanoparticles.

75 FR 22868 - Withdrawal of Regulatory Guide

Federal Register 2010, 2011, 2012, 2013, 2014

2010-04-30

... Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motion.'' FOR FURTHER.... Nuclear Regulatory Commission (NRC) is withdrawing RG 1.165, ``Identification and Characterization of... alter the licensing basis of any currently operating reactor or any of the currently issued early site...

Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

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

Ortensi, Javier; Baker, Benjamin Allen; Schunert, Sebastian

The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition,more » this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.« less

Minutes of the third annual meeting of the Panel on Reference Nuclear Data. [BNL, October 5, 1978

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

Burrows, T.W.; Stewart, L.; Coyne, J.J.

1979-05-01

The major activities of the meeting were as follows: welcome; organization, approval of minutes of the second meeting, and approval of agenda; review of nuclear data compilation and evaluation efforts (national and international efforts, master data files, publications); summary of 1977 panel meeting; definition of reference nuclear data; discussion of specific data needs and possible data center contributions (reactor physics, medicine and biology, controlled thermonuclear reactors and astrophysics); establishment of current interest and future direction of the panel; adjournment. Recommendations and action items are listed. Tables on nuclear data needs in applied physics, medicine and biology, and controlled thermonuclear reactorsmore » and astrophysics are presented. Appendixes include membership lists of various committees, summaries of publication activities, survey results, correspondence, and portions of the documents Proceedings of the Magnetic Fusion Energy Blanket and Shield Workshop and National Needs for Critically Evaluated Physical and Chemical Data. (RWR)« less

Engine management during NTRE start up

NASA Technical Reports Server (NTRS)

Bulman, Mel; Saltzman, Dave

1993-01-01

The topics are presented in viewgraph form and include the following: total engine system management critical to successful nuclear thermal rocket engine (NTRE) start up; NERVA type engine start windows; reactor power control; heterogeneous reactor cooling; propellant feed system dynamics; integrated NTRE start sequence; moderator cooling loop and efficient NTRE starting; analytical simulation and low risk engine development; accurate simulation through dynamic coupling of physical processes; and integrated NTRE and mission performance.

Production of bio-oil from underutilized forest biomass using an auger reactor

Treesearch

H. Ravindran; S. Thangalzhy-Gopakumar; S. Adhikari; O. Fasina; M. Tu; B. Via; E. Carter; S. Taylor

2015-01-01

Conversion of underutilized forest biomass to bio-oil could be a niche market for energy production. In this work, bio-oil was produced from underutilized forest biomass at selected temperatures between 425–500°C using an auger reactor. Physical properties of bio-oil, such as pH, density, heating value, ash, and water, were analyzed and compared with an ASTM standard...

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

Supplemental Reactor Physics Calculations and Analysis of ELF Mk 1A Fuel

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

Pope, Michael A.

2014-10-01

These calculations supplement previous the reactor physics work evaluating the Enhanced Low Enriched Uranium (LEU) Fuel (ELF) Mk 1A element. This includes various additional comparisons between the current Highly Enriched Uranium (HEU) and LEU along with further characterization of the performance of the ELF fuel. The excess reactivity to be held down at BOC for ELF Mk 1A fuel is estimated to be approximately $2.75 greater than with HEU for a typical cycle. This is a combined effect of the absence of burnable poison in the ELF fuel and the reduced neck shim worth in LEU fuel compared to HEU.more » Burnable poison rods were conceptualized for use in the small B positions containing Gd2O3 absorber. These were shown to provide $2.37 of negative reactivity at BOC and to burn out in less than half of a cycle. The worth of OSCCs is approximately the same between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. This was evaluated by rotating all banks simultaneously. The safety rod worth is relatively unchanged between HEU and ELF Mk 1A (LEU) fuels in the representative loading evaluated. However, this should be reevaluated with different loadings. Neutron flux, both total and fast (>1 MeV), is either the same or reduced upon changing from HEU to ELF Mk 1A (LEU) fuels in the representative loading evaluated. This is consistent with the well-established trend of lower neutron fluxes for a given power in LEU than HEU.The IPT loop void reactivity is approximately the same or less positive with ELF Mk 1A (LEU) fuel than HEU in the representative loading evaluated.« less

Kinetics of thermal decomposition of hydrated minerals associated with hematite ore in a fluidized bed reactor

NASA Astrophysics Data System (ADS)

Beuria, P. C.; Biswal, S. K.; Mishra, B. K.; Roy, G. G.

2017-03-01

The kinetics of removal of loss on ignition (LOI) by thermal decomposition of hydrated minerals present in natural iron ores (i.e., kaolinite, gibbsite, and goethite) was investigated in a laboratory-scale vertical fluidized bed reactor (FBR) using isothermal methods of kinetic analysis. Experiments in the FBR in batch processes were carried out at different temperatures (300 to 1200°C) and residence time (1 to 30 min) for four different iron ore samples with various LOIs (2.34wt% to 9.83wt%). The operating velocity was maintained in the range from 1.2 to 1.4 times the minimum fluidization velocity ( U mf). We observed that, below a certain critical temperature, the FBR did not effectively reduce the LOI to a desired level even with increased residence time. The results of this study indicate that the LOI level could be reduced by 90% within 1 min of residence time at 1100°C. The kinetics for low-LOI samples (<6wt%) indicates two different reaction mechanisms in two temperature regimes. At lower temperatures (300 to 700°C), the kinetics is characterized by a lower activation energy (diffusion-controlled physical moisture removal), followed by a higher activation energy (chemically controlled removal of LOI). In the case of high-LOI samples, three different kinetics mechanisms prevail at different temperature regimes. At temperature up to 450°C, diffusion kinetics prevails (removal of physical moisture); at temperature from 450 to 650°C, chemical kinetics dominates during removal of matrix moisture. At temperatures greater than 650°C, nucleation and growth begins to influence the rate of removal of LOI.

Grey water treatment in UASB reactor at ambient temperature.

PubMed

Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

2007-01-01

In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

Long Distance Reactor Antineutrino Flux Monitoring

NASA Astrophysics Data System (ADS)

Dazeley, Steven; Bergevin, Marc; Bernstein, Adam

2015-10-01

The feasibility of antineutrino detection as an unambiguous and unshieldable way to detect the presence of distant nuclear reactors has been studied. While KamLAND provided a proof of concept for long distance antineutrino detection, the feasibility of detecting single reactors at distances greater than 100 km has not yet been established. Even larger detectors than KamLAND would be required for such a project. Considerations such as light attenuation, environmental impact and cost, which favor water as a detection medium, become more important as detectors get larger. We have studied both the sensitivity of water based detection media as a monitoring tool, and the scientific impact such detectors might provide. A next generation water based detector may be able to contribute to important questions in neutrino physics, such as supernova neutrinos, sterile neutrino oscillations, and non standard electroweak interactions (using a nearby compact accelerator source), while also providing a highly sensitive, and inherently unshieldable reactor monitoring tool to the non proliferation community. In this talk I will present the predicted performance of an experimental non proliferation and high-energy physics program. Lawrence Livermore National Laboratory is operated by Lawrence Livermore National Security, LLC, for the U.S. Department of Energy, National Nuclear Security Administration under Contract DE-AC52-07NA27344. Release number LLNL-ABS-674192.

Benchmark Evaluation of the HTR-PROTEUS Absorber Rod Worths (Core 4)

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

John D. Bess; Leland M. Montierth

2014-06-01

PROTEUS was a zero-power research reactor at the Paul Scherrer Institute (PSI) in Switzerland. The critical assembly was constructed from a large graphite annulus surrounding a central cylindrical cavity. Various experimental programs were investigated in PROTEUS; during the years 1992 through 1996, it was configured as a pebble-bed reactor and designated HTR-PROTEUS. Various critical configurations were assembled with each accompanied by an assortment of reactor physics experiments including differential and integral absorber rod measurements, kinetics, reaction rate distributions, water ingress effects, and small sample reactivity effects [1]. Four benchmark reports were previously prepared and included in the March 2013 editionmore » of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) [2] evaluating eleven critical configurations. A summary of that effort was previously provided [3] and an analysis of absorber rod worth measurements for Cores 9 and 10 have been performed prior to this analysis and included in PROTEUS-GCR-EXP-004 [4]. In the current benchmark effort, absorber rod worths measured for Core Configuration 4, which was the only core with a randomly-packed pebble loading, have been evaluated for inclusion as a revision to the HTR-PROTEUS benchmark report PROTEUS-GCR-EXP-002.« less

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

Comparison of two modified coal ash ferric-carbon micro-electrolysis ceramic media for pretreatment of tetracycline wastewater.

PubMed

Yang, Kunlun; Jin, Yang; Yue, Qinyan; Zhao, Pin; Gao, Yuan; Wu, Suqing; Gao, Baoyu

2017-05-01

Application of modified sintering ferric-carbon ceramics (SFC) and sintering-free ferric-carbon ceramics (SFFC) based on coal ash and scrap iron for pretreatment of tetracycline (TET) wastewater was investigated in this article. Physical property, morphological character, toxic metal leaching content, and crystal component were studied to explore the application possibility of novel ceramics in micro-electrolysis reactors. The influences of operating conditions including influent pH, hydraulic retention time (HRT), and air-water ratio (A/W) on the removal of tetracycline were studied. The results showed that SFC and SFFC were suitable for application in micro-electrolysis reactors. The optimum conditions of SFC reactor were pH of 3, HRT of 7 h, and A/W of 10. For SFFC reactor, the optimum conditions were pH of 2, HRT of 7 h, and A/W of 15. In general, the TET removal efficiency of SFC reactor was better than that of SFFC reactor. However, the harden resistance of SFFC was better than that of SFC. Furthermore, the biodegradability of TET wastewater was improved greatly after micro-electrolysis pretreatment for both SFC and SFFC reactors.

Flow reversal power limit for the HFBR

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

Cheng, L.Y.; Tichler, P.R.

The High Flux Beam Reactor (HFBR) is a pressurized heavy water moderated and cooled research reactor that began operation at 40 MW. The reactor was subsequently upgraded to 60 MW and operated at that level for several years. The reactor undergoes a buoyancy-driven reversal of flow in the reactor core following certain postulated accidents. Questions which were raised about the afterheat removal capability during the flow reversal transition led to a reactor shutdown and subsequent resumption of operation at a reduced power of 30 MW. An experimental and analytical program to address these questions is described in this report. Themore » experiments were single channel flow reversal tests under a range of conditions. The analytical phase involved simulations of the tests to benchmark the physical models and development of a criterion for dryout. The criterion is then used in simulations of reactor accidents to determine a safe operating power level. It is concluded that the limit on the HFBR operating power with respect to the issue of flow reversal is in excess of 60 MW. Direct use of the experimental results and an understanding of the governing phenomenology supports this conclusion.« less

Use of high-volume outdoor smog chamber photo-reactors for studying physical and chemical atmospheric aerosol formation and composition

NASA Astrophysics Data System (ADS)

Borrás, E.; Ródenas, M.; Vera, T.; Muñoz, A.

2015-12-01

The atmospheric particulate matter has a large impact on climate, biosphere behaviour and human health. Its study is complex because of large number of species are present at low concentrations and the continuous time evolution, being not easily separable from meteorology, and transport processes. Closed systems have been proposed by isolating specific reactions, pollutants or products and controlling the oxidizing environment. High volume simulation chambers, such as EUropean PHOtoREactor (EUPHORE), are an essential tool used to simulate atmospheric photochemical reactions. This communication describes the last results about the reactivity of prominent atmospheric pollutants and the subsequent particulate matter formation. Specific experiments focused on organic aerosols have been developed at the EUPHORE photo-reactor. The use of on-line instrumentation, supported by off-line techniques, has provided well-defined reaction profiles, physical properties, and up to 300 different species are determined in particulate matter. The application fields include the degradation of anthropogenic and biogenic pollutants, and pesticides under several atmospheric conditions, studying their contribution on the formation of secondary organic aerosols (SOA). The studies performed at the EUPHORE have improved the mechanistic studies of atmospheric degradation processes and the knowledge about the chemical and physical properties of atmospheric particulate matter formed during these processes.

Nondiffusive transport regimes for suprathermal ions in turbulent plasmas

NASA Astrophysics Data System (ADS)

Bovet, A.; Fasoli, A.; Ricci, P.; Furno, I.; Gustafson, K.

2015-04-01

The understanding of the transport of suprathermal ions in the presence of turbulence is important for fusion plasmas in the burning regime that will characterize reactors, and for space plasmas to understand the physics of particle acceleration. Here, three-dimensional measurements of a suprathermal ion beam in the toroidal plasma device TORPEX are presented. These measurements demonstrate, in a turbulent plasma, the existence of subdiffusive and superdiffusive transport of suprathermal ions, depending on their energy. This result stems from the unprecedented combination of uniquely resolved measurements and first-principles numerical simulations that reveal the mechanisms responsible for the nondiffusive transport. The transport regime is determined by the interaction of the suprathermal ion orbits with the turbulent plasma dynamics, and is strongly affected by the ratio of the suprathermal ion energy to the background plasma temperature.

PREFACE: SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor: Devoted to the 75th anniversary of Yu M Ostanevich's birth

NASA Astrophysics Data System (ADS)

Gordely, Valentin; Kuklin, Alexander; Balasoiu, Maria

2012-03-01

The Second International Workshop 'SANS-YuMO User Meeting at the Start-up of Scientific Experiments on the IBR-2M Reactor', devoted to the 75th anniversary of the birth of Professor Yu M Ostanevich (1936-1992), an outstanding neutron physicist and the founder of small-angle neutron scattering (field, group, and instrument) at JINR FLNPh, was held on 27-30 May at the Frank Laboratory of Neutron Physics. The first Workshop was held in October 2006. Research groups from different neutron centers, universities and research institutes across Europe presented more than 35 oral and poster presentations describing scientific and methodological results. Most of them were obtained with the help of the YuMO instrument before the IBR-2 shutdown in 2006. For the last four years the IBR-2 reactor has been shut down for refurbishment. At the end of 2010 the physical launch of the IBR-2M reactor was finally realized. Nowadays the small-angle neutron scattering (SANS) technique is applied to a wide range of scientific problems in condensed matter, soft condensed matter, biology and nanotechnology, and despite the fact that there are currently over 30 SANS instruments in operation worldwide at both reactor and spallation sources, the demand for beam-time is considerably higher than the time available. It must be remembered, however, that as the first SANS machine on a steady-state reactor was constructed at the Institute Laue Langevin, Grenoble, the first SANS instrument on a 'white' neutron pulsed beam was accomplished at the Joint Institute for Nuclear Research at the IBR-30 reactor, beamline N5. During the meeting Yu M Ostanevich's determinative and crucial contribution to the construction of spectrometers at the IBR-2 high-pulsed reactor was presented, as well as his contribution to the development of the time-of-flight (TOF) small-angle scattering technique, and a selection of other scientific areas. His leadership and outstanding scientific achievements in applications of the Mossbauer effect in physics and chemistry, in SANS studies of polyelectrolytes, small molecules, fractals, metallic glasses, macromolecules, polymers, etc., were recognized by a number of awards including the State Prize of the Russian Federation in 2000. The scientific program of the workshop focused on fundamental and methodical research at the YuMO spectrometer and developments of the SANS instrument at the modernized IBR-2M reactor. We recall that the acronym YuMO of the small-angle neutron scattering spectrometer (MURN), was given in honor of Yu M Ostanevich. One of the most important objectives of this user meeting was to discuss the further development possibilities of the YuMO spectrometer with experts, in the frame of a SANS YuMO Round Table, taking into account the specific performance of the modernized YuMO SANS instrument, and the scientific and technical requests of the instrument's users. Highlights on modern achievements in nanoscience, polymers and biology were other significant goals of the meeting. The plenary invited talks were presented by leading scientists in small-angle neutron scattering and soft condensed matter, including members of the Russian Academy of Sciences: Prof. Heinrich Stuhrmann, Prof. Alexei Khokhlov, Prof. Jose Teixeira, Prof. Alexander Ozerin, Prof. Albrecht Wiedenmann, etc. There were 27 oral talks given and 32 posters presented by 92 participants from 12 countries: Czech Republic, Egypt, France, Germany, Hungary, Moldova, Mongolia, Poland, Romania, Russian Federation, Slovak Republic, and Ukraine. The workshop was organized with the financial support of the Frank Laboratory of Neutron Physics (Joint Institute for Nuclear Research), Horia Hulubei National Institute of Physics and Nuclear Engineering - IFIN HH (Romania), Institute of Macromolecular Chemistry AS CR (Czech Republic), and Comenius University (Slovakia). V Gordeliy, A Kuklin and M Balasoiu SANSgroup Participants of the meeting The PDF also contains additional photographs from the meeting.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

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

Reactor physics phenomena in additively manufactured control elements for the High Flux Isotope Reactor

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

Burns, Joseph R.; Petrovic, Bojan; Chandler, David

Additive manufacturing is under investigation as a novel method of fabricating the control elements (CEs) of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory with greater simplicity, eliminating numerous highly complex fabrication steps and thereby offering potential for significant savings in cost, time, and effort. This process yields a unique CE design with lumped absorbers, a departure from traditionally manufactured CEs with uniformly distributed absorbing material. Here, this study undertakes a neutronics analysis of the impact of additively manufactured CEs on the HFIR core physics, seeking preliminary assessment of the feasibility of their practical use. The resultsmore » of the MCNP transport simulations reveal changes in the HFIR reactor physics arising from geometric and nuclear effects. Absorber lumping in the discrete CEs yields a large volume of unpoisoned material that is not present in the homogeneous design, in turn yielding increases in free thermal flux in the CE absorbing regions and their immediate vicinity. The availability of additional free thermal neutrons in the core yields an increase in fission rate density in the fuel closest to the CEs and a corresponding increase in neutron multiplication on the order of 100 pcm. The absorption behavior exhibited by the discrete CEs is markedly different from the homogeneous CEs due to several competing effects. Self-shielding arising from absorber lumping acts to reduce the effective absorption cross section of the discrete CEs, but this effect is offset by geometric and spectral effects. The operational performance of the discrete CEs is found to be comparable to the homogeneous CEs, with only limited deficiencies in reactivity worth that are expected to be operationally recoverable via limited adjustment of the CE positions and withdrawal rate. On the whole, these results indicate that the discrete CEs perform reasonably similarly to the homogeneous CEs and appear feasible for application in HFIR. In conclusion, the physical phenomena identified in this study provide valuable background for follow-up design studies.« less

Reactor physics phenomena in additively manufactured control elements for the High Flux Isotope Reactor

DOE PAGES

Burns, Joseph R.; Petrovic, Bojan; Chandler, David; ...

2018-02-22

Additive manufacturing is under investigation as a novel method of fabricating the control elements (CEs) of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory with greater simplicity, eliminating numerous highly complex fabrication steps and thereby offering potential for significant savings in cost, time, and effort. This process yields a unique CE design with lumped absorbers, a departure from traditionally manufactured CEs with uniformly distributed absorbing material. Here, this study undertakes a neutronics analysis of the impact of additively manufactured CEs on the HFIR core physics, seeking preliminary assessment of the feasibility of their practical use. The resultsmore » of the MCNP transport simulations reveal changes in the HFIR reactor physics arising from geometric and nuclear effects. Absorber lumping in the discrete CEs yields a large volume of unpoisoned material that is not present in the homogeneous design, in turn yielding increases in free thermal flux in the CE absorbing regions and their immediate vicinity. The availability of additional free thermal neutrons in the core yields an increase in fission rate density in the fuel closest to the CEs and a corresponding increase in neutron multiplication on the order of 100 pcm. The absorption behavior exhibited by the discrete CEs is markedly different from the homogeneous CEs due to several competing effects. Self-shielding arising from absorber lumping acts to reduce the effective absorption cross section of the discrete CEs, but this effect is offset by geometric and spectral effects. The operational performance of the discrete CEs is found to be comparable to the homogeneous CEs, with only limited deficiencies in reactivity worth that are expected to be operationally recoverable via limited adjustment of the CE positions and withdrawal rate. On the whole, these results indicate that the discrete CEs perform reasonably similarly to the homogeneous CEs and appear feasible for application in HFIR. In conclusion, the physical phenomena identified in this study provide valuable background for follow-up design studies.« less

Integral Full Core Multi-Physics PWR Benchmark with Measured Data

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

Forget, Benoit; Smith, Kord; Kumar, Shikhar

In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevantmore » multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.« less

Heat Source Characterization In A TREAT Fuel Particle Using Coupled Neutronics Binary Collision Monte-Carlo Calculations

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

Schunert, Sebastian; Schwen, Daniel; Ghassemi, Pedram

This work presents a multi-physics, multi-scale approach to modeling the Transient Test Reactor (TREAT) currently prepared for restart at the Idaho National Laboratory. TREAT fuel is made up of microscopic fuel grains (r ˜ 20µm) dispersed in a graphite matrix. The novelty of this work is in coupling a binary collision Monte-Carlo (BCMC) model to the Finite Element based code Moose for solving a microsopic heat-conduction problem whose driving source is provided by the BCMC model tracking fission fragment energy deposition. This microscopic model is driven by a transient, engineering scale neutronics model coupled to an adiabatic heating model. Themore » macroscopic model provides local power densities and neutron energy spectra to the microscpic model. Currently, no feedback from the microscopic to the macroscopic model is considered. TREAT transient 15 is used to exemplify the capabilities of the multi-physics, multi-scale model, and it is found that the average fuel grain temperature differs from the average graphite temperature by 80 K despite the low-power transient. The large temperature difference has strong implications on the Doppler feedback a potential LEU TREAT core would see, and it underpins the need for multi-physics, multi-scale modeling of a TREAT LEU core.« less

Advances in modelling of condensation phenomena

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

Liu, W.S.; Zaltsgendler, E.; Hanna, B.

1997-07-01

The physical parameters in the modelling of condensation phenomena in the CANDU reactor system codes are discussed. The experimental programs used for thermal-hydraulic code validation in the Canadian nuclear industry are briefly described. The modelling of vapour generation and in particular condensation plays a key role in modelling of postulated reactor transients. The condensation models adopted in the current state-of-the-art two-fluid CANDU reactor thermal-hydraulic system codes (CATHENA and TUF) are described. As examples of the modelling challenges faced, the simulation of a cold water injection experiment by CATHENA and the simulation of a condensation induced water hammer experiment by TUFmore » are described.« less

Physics Teachers Workshop

ScienceCinema

Huggins, DaNel; Calhoun, John; Palmer, Alyson; Thorpe, Steve; Vanderveen, Anne

2017-12-09

INL is looking for the nation's top high school physics teachers to attend our July workshop in Idaho Falls. Participants get to learn from nuclear researchers, tour facilities including a research reactor and interact with peers from across the country. You can learn more about INL projects at http://www.facebook.com/idahonationallaboratory

Physics Teachers Workshop

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

Huggins, DaNel; Calhoun, John; Palmer, Alyson

INL is looking for the nation's top high school physics teachers to attend our July workshop in Idaho Falls. Participants get to learn from nuclear researchers, tour facilities including a research reactor and interact with peers from across the country. You can learn more about INL projects at http://www.facebook.com/idahonationallaboratory

EDITORIAL: Plasma Surface Interactions for Fusion

NASA Astrophysics Data System (ADS)

2006-05-01

Because plasma-boundary physics encompasses some of the most important unresolved issues for both the International Thermonuclear Experimental Reactor (ITER) project and future fusion power reactors, there is a strong interest in the fusion community for better understanding and characterization of plasma wall interactions. Chemical and physical sputtering cause the erosion of the limiters/divertor plates and vacuum vessel walls (made of C, Be and W, for example) and degrade fusion performance by diluting the fusion fuel and excessively cooling the core, while carbon redeposition could produce long-term in-vessel tritium retention, degrading the superior thermo-mechanical properties of the carbon materials. Mixed plasma-facing materials are proposed, requiring optimization for different power and particle flux characteristics. Knowledge of material properties as well as characteristics of the plasma material interaction are prerequisites for such optimizations. Computational power will soon reach hundreds of teraflops, so that theoretical and plasma science expertise can be matched with new experimental capabilities in order to mount a strong response to these challenges. To begin to address such questions, a Workshop on New Directions for Advanced Computer Simulations and Experiments in Fusion-Related Plasma Surface Interactions for Fusion (PSIF) was held at the Oak Ridge National Laboratory from 21 to 23 March, 2005. The purpose of the workshop was to bring together researchers in fusion related plasma wall interactions in order to address these topics and to identify the most needed and promising directions for study, to exchange opinions on the present depth of knowledge of surface properties for the main fusion-related materials, e.g., C, Be and W, especially for sputtering, reflection, and deuterium (tritium) retention properties. The goal was to suggest the most important next steps needed for such basic computational and experimental work to be facilitated by researchers in fusion, material, and physical sciences. Representatives from many fusion research laboratories attended, and 25 talks were given, the majority of them making up the content of these Workshop proceedings. The presentations of all talks and further information on the Workshop are available at http://www-cfadc.phy.ornl.gov/psif/home.html. The workshop talks dealt with identification of needs from the perspective of integrated fusion simulation and ITER design, recent developments and perspectives on computation of plasma-facing surface properties using the current and expected new generation of computation capability, and with the status of dedicated laboratory experiments which characterize the underlying processes of PSIF. The Workshop summary and conclusions are being published in Nuclear Fusion 45 (2005). We are indebted to Lynda Saddiq and Fay Ownby, secretaries in the Physics Division of ORNL, whose special efforts, devotion, and expertise made possible both the Workshop and these Proceedings. J T Hogan, P S Krstic and F W Meyer Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6372, USA

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

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

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

Vegetation survey of PEN Branch wetlands

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

Not Available

A survey was conducted of vegetation along Pen Branch Creek at Savannah River Site (SRS) in support of K-Reactor restart. Plants were identified to species by overstory, understory, shrub, and groundcover strata. Abundance was also characterized and richness and diversity calculated. Based on woody species basal area, the Pen Branch delta was the most impacted, followed by the sections between the reactor and the delta. Species richness for shrub and groundcover strata were also lowest in the delta. No endangered plant species were found. Three upland pine areas were also sampled. In support of K Reactor restart, this report summarizesmore » a study of the wetland vegetation along Pen Branch. Reactor effluent enters Indian Grove Branch and then flows into Pen Branch and the Pen Branch Delta.« less

Vegetation survey of PEN Branch wetlands

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

Not Available

1991-01-01

A survey was conducted of vegetation along Pen Branch Creek at Savannah River Site (SRS) in support of K-Reactor restart. Plants were identified to species by overstory, understory, shrub, and groundcover strata. Abundance was also characterized and richness and diversity calculated. Based on woody species basal area, the Pen Branch delta was the most impacted, followed by the sections between the reactor and the delta. Species richness for shrub and groundcover strata were also lowest in the delta. No endangered plant species were found. Three upland pine areas were also sampled. In support of K Reactor restart, this report summarizesmore » a study of the wetland vegetation along Pen Branch. Reactor effluent enters Indian Grove Branch and then flows into Pen Branch and the Pen Branch Delta.« less

Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

NASA Astrophysics Data System (ADS)

Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

2015-09-01

Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

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

Unwin, Stephen D.; Lowry, Peter P.; Layton, Robert F.

This is a working report drafted under the Risk-Informed Safety Margin Characterization pathway of the Light Water Reactor Sustainability Program, describing statistical models of passives component reliabilities.

Lithographically fabricated silicon microreactor for in situ characterization of heterogeneous catalysts—Enabling correlative characterization techniques

NASA Astrophysics Data System (ADS)

Baier, S.; Rochet, A.; Hofmann, G.; Kraut, M.; Grunwaldt, J.-D.

2015-06-01

We report on a new modular setup on a silicon-based microreactor designed for correlative spectroscopic, scattering, and analytic on-line gas investigations for in situ studies of heterogeneous catalysts. The silicon microreactor allows a combination of synchrotron radiation based techniques (e.g., X-ray diffraction and X-ray absorption spectroscopy) as well as infrared thermography and Raman spectroscopy. Catalytic performance can be determined simultaneously by on-line product analysis using mass spectrometry. We present the design of the reactor, the experimental setup, and as a first example for an in situ study, the catalytic partial oxidation of methane showing the applicability of this reactor for in situ studies.

Lithographically fabricated silicon microreactor for in situ characterization of heterogeneous catalysts—Enabling correlative characterization techniques.

PubMed

Baier, S; Rochet, A; Hofmann, G; Kraut, M; Grunwaldt, J-D

2015-06-01

We report on a new modular setup on a silicon-based microreactor designed for correlative spectroscopic, scattering, and analytic on-line gas investigations for in situ studies of heterogeneous catalysts. The silicon microreactor allows a combination of synchrotron radiation based techniques (e.g., X-ray diffraction and X-ray absorption spectroscopy) as well as infrared thermography and Raman spectroscopy. Catalytic performance can be determined simultaneously by on-line product analysis using mass spectrometry. We present the design of the reactor, the experimental setup, and as a first example for an in situ study, the catalytic partial oxidation of methane showing the applicability of this reactor for in situ studies.

Determining the morphology of polystyrene-block-poly(2-vinylpyridine) micellar reactors for ZnO nanoparticle synthesis.

PubMed

El-Atwani, Osman; El-Atwani, Osman C; Aytun, Taner; Mutaf, Omer Faruk; Srot, Vesna; van Aken, Peter A; Ow-Yang, Cleva W

2010-05-18

We report the use of reverse PS-b-P2VP diblock copolymer micelles as true nanoscale-sized reactor vessels to synthesize ZnO nanoparticles. The reverse micelles were formed in toluene and then sequentially loaded with zinc acetate dihydrate and tetramethylammonium hydroxide reactants. Moreover, high spatial resolution Z-contrast imaging and EDX spectroscopy techniques were used to confirm the segregation of the Zn cation to the core of the loaded micelles. Determining the chemical distribution with high nanoscale spatial resolution is shown to complement the less direct characterization by AFM, DLS and FTIR, thus demonstrating broader implications for the characterization of hybrid nanocomposite systems.

Characterization of undissolved solids from the dissolution of North Anna reactor fuel

DOE PAGES

Rudisill, Tracy S.; Olson, L. C.; DiPrete, D. P.

2017-06-16

Here, samples of undissolved solids (UDS) from the dissolution of North Anna reactor fuel were characterized to investigate the effects of using air or oxygen as the oxidant during tritium removal. The UDS composition data also support the development of a waste form for disposal. There was no discernible effect of the oxidant used during the tritium removal process or the size fraction on the UDS composition. Scanning electron microscopy (SEM) and energy dispersive (x-ray) spectroscopy were used to estimate the oxygen content of the UDS and it was found to be potentially significant, on the order of 30% bymore » mass and 80% by atom.« less

Can high fields save the tokamak? The challenge of steady-state operation for low cost compact reactors

NASA Astrophysics Data System (ADS)

Freidberg, Jeffrey; Dogra, Akshunna; Redman, William; Cerfon, Antoine

2016-10-01

The development of high field, high temperature superconductors is thought to be a game changer for the development of fusion power based on the tokamak concept. We test the validity of this assertion for pilot plant scale reactors (Q 10) for two different but related missions: pulsed operation and steady-state operation. Specifically, we derive a set of analytic criteria that determines the basic design parameters of a given fusion reactor mission. As expected there are far more constraints than degrees of freedom in any given design application. However, by defining the mission of the reactor under consideration, we have been able to determine the subset of constraints that drive the design, and calculate the values for the key parameters characterizing the tokamak. Our conclusions are as follows: 1) for pulsed reactors, high field leads to more compact designs and thus cheaper reactors - high B is the way to go; 2) steady-state reactors with H-mode like transport are large, even with high fields. The steady-state constraint is hard to satisfy in compact designs - high B helps but is not enough; 3) I-mode like transport, when combined with high fields, yields relatively compact steady-state reactors - why is there not more research on this favorable transport regime?

CESAR5.3: Isotopic depletion for Research and Testing Reactor decommissioning

NASA Astrophysics Data System (ADS)

Ritter, Guillaume; Eschbach, Romain; Girieud, Richard; Soulard, Maxime

2018-05-01

CESAR stands in French for "simplified depletion applied to reprocessing". The current version is now number 5.3 as it started 30 years ago from a long lasting cooperation with ORANO, co-owner of the code with CEA. This computer code can characterize several types of nuclear fuel assemblies, from the most regular PWR power plants to the most unexpected gas cooled and graphite moderated old timer research facility. Each type of fuel can also include numerous ranges of compositions like UOX, MOX, LEU or HEU. Such versatility comes from a broad catalog of cross section libraries, each corresponding to a specific reactor and fuel matrix design. CESAR goes beyond fuel characterization and can also provide an evaluation of structural materials activation. The cross-sections libraries are generated using the most refined assembly or core level transport code calculation schemes (CEA APOLLO2 or ERANOS), based on the European JEFF3.1.1 nuclear data base. Each new CESAR self shielded cross section library benefits all most recent CEA recommendations as for deterministic physics options. Resulting cross sections are organized as a function of burn up and initial fuel enrichment which allows to condensate this costly process into a series of Legendre polynomials. The final outcome is a fast, accurate and compact CESAR cross section library. Each library is fully validated, against a stochastic transport code (CEA TRIPOLI 4) if needed and against a reference depletion code (CEA DARWIN). Using CESAR does not require any of the neutron physics expertise implemented into cross section libraries generation. It is based on top quality nuclear data (JEFF3.1.1 for ˜400 isotopes) and includes up to date Bateman equation solving algorithms. However, defining a CESAR computation case can be very straightforward. Most results are only 3 steps away from any beginner's ambition: Initial composition, in core depletion and pool decay scenario. On top of a simple utilization architecture, CESAR includes a portable Graphical User Interface which can be broadly deployed in R&D or industrial facilities. Aging facilities currently face decommissioning and dismantling issues. This way to the end of the nuclear fuel cycle requires a careful assessment of source terms in the fuel, core structures and all parts of a facility that must be disposed of with "industrial nuclear" constraints. In that perspective, several CESAR cross section libraries were constructed for early CEA Research and Testing Reactors (RTR's). The aim of this paper is to describe how CESAR operates and how it can be used to help these facilities care for waste disposal, nuclear materials transport or basic safety cases. The test case will be based on the PHEBUS Facility located at CEA - Cadarache.

NUCLEAR REACTOR

DOEpatents

Grebe, J.J.

1959-07-14

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

Synthesis and Manipulation of Semiconductor Nanocrystals inMicrofluidic Reactors

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

Chan, Emory Ming-Yue

2006-01-01

Microfluidic reactors are investigated as a mechanism tocontrol the growth of semiconductor nanocrystals and characterize thestructural evolution of colloidal quantum dots. Due to their shortdiffusion lengths, low thermal masses, and predictable fluid dynamics,microfluidic devices can be used to quickly and reproducibly alterreaction conditions such as concentration, temperature, and reactiontime, while allowing for rapid reagent mixing and productcharacterization. These features are particularly useful for colloidalnanocrystal reactions, which scale poorly and are difficult to controland characterize in bulk fluids. To demonstrate the capabilities ofnanoparticle microreactors, a size series of spherical CdSe nanocrystalswas synthesized at high temperature in a continuous-flow, microfabricatedglass reactor. Nanocrystalmore » diameters are reproducibly controlled bysystematically altering reaction parameters such as the temperature,concentration, and reaction time. Microreactors with finer control overtemperature and reagent mixing were designed to synthesize nanoparticlesof different shapes, such as rods, tetrapods, and hollow shells. The twomajor challenges observed with continuous flow reactors are thedeposition of particles on channel walls and the broad distribution ofresidence times that result from laminar flow. To alleviate theseproblems, I designed and fabricated liquid-liquid segmented flowmicroreactors in which the reaction precursors are encapsulated inflowing droplets suspended in an immiscible carrier fluid. The synthesisof CdSe nanocrystals in such microreactors exhibited reduced depositionand residence time distributions while enabling the rapid screening aseries of samples isolated in nL droplets. Microfluidic reactors werealso designed to modify the composition of existing nanocrystals andcharacterize the kinetics of such reactions. The millisecond kinetics ofthe CdSe-to-Ag 2Se nanocrystal cation exchange reaction are measured insitu with micro-X-ray Absorption Spectroscopy in silicon microreactorsspecifically designed for rapid mixing and time-resolved X-rayspectroscopy. These results demonstrate that microreactors are valuablefor controlling and characterizing a wide range of reactions in nLvolumes even when nanoscale particles, high temperatures, causticreagents, and rapid time scales are involved. These experiments providethe foundation for future microfluidic investigations into the mechanismsof nanocrystal growth, crystal phase evolution, and heterostructureassembly.« less

New results from RENO and prospects with RENO-50

NASA Astrophysics Data System (ADS)

Kim, Soo-Bong

2015-08-01

RENO (Reactor Experiment for Neutrino Oscillation) has made a definitive measurement of the smallest mixing angle θ13 in 2012, based on the disappearance of electron antineutrinos. More precise measurements have been obtained and presented on the mixing angle and the reactor neutrino spectrum, using ˜800 days of data. We have observed an excess of inverse-beta-decay (IBD) prompt spectrum near 5 MeV with respect to the most commonly used models. The excess is found to be consistent with coming from reactors. We have also successfully measured the reactor neutrino flux with a delayed signal of neutron capture on hydrogen. A future reactor neutrino experiment of RENO-50 is proposed to determine the neutrino mass hierarchy and to make highly precise measurements of θ12, Δm212, and Δ m312. Physics goals and sensitivities of RENO-50 are presented with a strategy of obtaining an extremely good energy resolution toward the neutrino mass hierarchy.

Space rocket engine on the base of the reactor-pumped laser for the interplanetary flights and earth orbital applications

NASA Astrophysics Data System (ADS)

Gulevich, Andrey V.; Dyachenko, Peter P.; Kukharchuk, Oleg F.; Zrodnikov, Anatoly V.

2000-01-01

In this report the concept of vehicle-based reactor-laser engine for long time interplanetary and interorbital (LEO to GEO) flights is proposed. Reactor-pumped lasers offer the perspective way to create on the base of modern nuclear and lasers technologies the low mass and high energy density, repetitively pulsed vehicle-based laser of average power 100 kW. Nowadays the efficiency of nuclear-to-optical energy conversion reached the value of 2-3%. The demo model of reactor-pumped laser facility is under construction in Institute for Physics and Power Engineering (Obninsk, Russia). It enable us to hope that using high power laser on board of the vehicle could make the effective space laser engine possible. Such engine may provide the high specific impulse ~1000-2000 s with the thrust up to 10-100 n. Some calculation results of the characteristics of vehicle-based reactor-laser thermal engine concept are also presented. .

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

Divertor detachment

NASA Astrophysics Data System (ADS)

Krasheninnikov, Sergei

2015-11-01

The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

Self-potential and Complex Conductivity Monitoring of In Situ Hydrocarbon Remediation in Microbial Fuel Cell

NASA Astrophysics Data System (ADS)

Zhang, C.; Revil, A.; Ren, Z.; Karaoulis, M.; Mendonca, C. A.

2013-12-01

Petroleum hydrocarbon contamination of soil and groundwater in both non-aqueous phase liquid and dissolved forms generated from spills and leaks is a wide spread environmental issue. Traditional cleanup of hydrocarbon contamination in soils and ground water using physical, chemical, and biological remedial techniques is often expensive and ineffective. Recent studies show that the microbial fuel cell (MFC) can simultaneously enhance biodegradation of hydrocarbons in soil and groundwater and yield electricity. Non-invasive geophysical techniques such as self-potential (SP) and complex conductivity (induced polarization) have shown the potential to detect and characterize the nature of electron transport mechanism of in situ bioremediation of organic contamination plumes. In this study, we deployed both SP and complex conductivity in lab scale MFCs to monitor time-laps geophysical response of degradation of hydrocarbons by MFC. Two different sizes of MFC reactors were used in this study (DI=15 cm cylinder reactor and 94.5cm x 43.5 cm rectangle reactor), and the initial hydrocarbon concentration is 15 g diesel/kg soil. SP and complex conductivity measurements were measured using non-polarizing Ag/AgCl electrodes. Sensitivity study was also performed using COMSOL Multiphysics to test different electrode configurations. The SP measurements showed stronger anomalies adjacent to the MFC than locations afar, and both real and imaginary parts of complex conductivity are greater in areas close to MFC than areas further away and control samples without MFC. The joint use of SP and complex conductivity could in situ evaluate the dynamic changes of electrochemical parameters during this bioremediation process at spatiotemporal scales unachievable with traditional sampling methods. The joint inversion of these two methods to evaluate the efficiency of MFC enhanced hydrocarbon remediation in the subsurface.

ETR, TRA642 AND TRA647. FLOOR PLANS FOR FIRST AND SECOND ...

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

ETR, TRA-642 AND TRA-647. FLOOR PLANS FOR FIRST AND SECOND FLOORS OF THE OFFICE AND CONTROL BUILDING ALONG THE NORTH WALL OF THE ETR BUILDING. HEALTH PHYSICS, OPERATIONS, AND CONTROL ROOM. AIRLOCK DOOR. OFFICES. STAIRWAY LOCATIONS. KAISER ETR-5528-MTR-642-A-3, 10/1955. INL INDEX NO. 532-0642-00-100911, REV. 0. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID