On some control problems of dynamic of reactor

NASA Astrophysics Data System (ADS)

Baskakov, A. V.; Volkov, N. P.

2017-12-01

The paper analyzes controllability of the transient processes in some problems of nuclear reactor dynamics. In this case, the mathematical model of nuclear reactor dynamics is described by a system of integro-differential equations consisting of the non-stationary anisotropic multi-velocity kinetic equation of neutron transport and the balance equation of delayed neutrons. The paper defines the formulation of the linear problem on control of transient processes in nuclear reactors with application of spatially distributed actions on internal neutron sources, and the formulation of the nonlinear problems on control of transient processes with application of spatially distributed actions on the neutron absorption coefficient and the neutron scattering indicatrix. The required control actions depend on the spatial and velocity coordinates. The theorems on existence and uniqueness of these control actions are proved in the paper. To do this, the control problems mentioned above are reduced to equivalent systems of integral equations. Existence and uniqueness of the solution for this system of integral equations is proved by the method of successive approximations, which makes it possible to construct an iterative scheme for numerical analyses of transient processes in a given nuclear reactor with application of the developed mathematical model. Sufficient conditions for controllability of transient processes are also obtained. In conclusion, a connection is made between the control problems and the observation problems, which, by to the given information, allow us to reconstruct either the function of internal neutron sources, or the neutron absorption coefficient, or the neutron scattering indicatrix....

Dislocations et propriétés mécaniques des matériaux céramiques: Quelques problèmes

NASA Astrophysics Data System (ADS)

Castaing, J.; Dominguez Rodriguez, A.

1995-11-01

The study of plastic deformation of ceramic materials raised new problems on low temperature dislocation glide and high temperature dislocation climb. Mechanical behaviour can be explained. In this paper, we review some examples related to oxides which are linked to the activity of J. Philibert. L'étude de la déformation plastique de matériaux céramiques monocristallins a donné l'occasion de poser des nouveaux problèmes sur le glissement des dislocations à basse température et sur leur montée à haute température. Le comportement mécanique peut ainsi être expliqué. Nous passons en revue des cas concernant les oxydes dans lesquels J. Philibert a joué un rôle important.

Transmutation of Isotopes --- Ecological and Energy Production Aspects

NASA Astrophysics Data System (ADS)

Gudowski, Waclaw

2000-01-01

This paper describes principles of Accelerator-Driven Transmutation of Nuclear Wastes (ATW) and gives some flavour of the most important topics which are today under investigations in many countries. An assessment of the potential impact of ATW on a future of nuclear energy is also given. Nuclear reactors based on self-sustained fission reactions --- after spectacular development in fifties and sixties, that resulted in deployment of over 400 power reactors --- are wrestling today more with public acceptance than with irresolvable technological problems. In a whole spectrum of reasons which resulted in today's opposition against nuclear power few of them are very relevant for the nuclear physics community and they arose from the fact that development of nuclear power had been handed over to the nuclear engineers and technicians with some generically unresolved problems, which should have been solved properly by nuclear scientists. In a certain degree of simplification one can say, that most of the problems originate from very specific features of a fission phenomenon: self-sustained chain reaction in fissile materials and very strong radioactivity of fission products and very long half-life of some of the fission and activation products. And just this enormous concentration of radioactive fission products in the reactor core is the main problem of managing nuclear reactors: it requires unconditional guarantee for the reactor core integrity in order to avoid radioactive contamination of the environment; it creates problems to handle decay heat in the reactor core and finally it makes handling and/or disposal of spent fuel almost a philosophical issue, due to unimaginable long time scales of radioactive decay of some isotopes. A lot can be done to improve the design of conventional nuclear reactors (like Light Water Reactors); new, better reactors can be designed but it seems today very improbable to expect any radical change in the public perception of conventional nuclear power. In this context a lot of hopes and expectations have been expressed for novel systems called Accelerator-Driven Systems, Accelerator-Driven Transmutation of Waste or just Hybrid Reactors. All these names are used for description of the same nuclear system combining a powerful particle accelerator with a subcritical reactor. A careful analysis of possible environmental impact of ATW together with limitation of this technology is presented also in this paper.

Quelques considerations sur la traduction medicale et pharmaceutique (Some Considerations in Medical and Pharmaceutical Translation)

ERIC Educational Resources Information Center

Sliosberg, A.

1971-01-01

Paper presented during the meeting of the Section Presse et Documentation" of the 29th International Congress of Pharmaceutical Science of the International Pharmaceutical Federation, London, September 10, 1969. (VM)

Comparative evaluation of solar, fission, fusion, and fossil energy resources, part 3

NASA Technical Reports Server (NTRS)

Clement, J. D.; Reupke, W. A.

1974-01-01

The role of nuclear fission reactors in becoming an important power source in the world is discussed. The supply of fissile nuclear fuel will be severely depleted by the year 2000. With breeder reactors the world supply of uranium could last thousands of years. However, breeder reactors have problems of a large radioactive inventory and an accident potential which could present an unacceptable hazard. Although breeder reactors afford a possible solution to the energy shortage, their ultimate role will depend on demonstrated safety and acceptable risks and environmental effects. Fusion power would also be a long range, essentially permanent, solution to the world's energy problem. Fusion appears to compare favorably with breeders in safety and environmental effects. Research comparing a controlled fusion reactor with the breeder reactor in solving our long range energy needs is discussed.

Graphite distortion ``C`` Reactor. Final report

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

Wood, N.H.

1962-02-08

This report covers the efforts of the Laboratory in an investigation of the graphite distortion in the ``C`` reactor at Hanford. The particular aspects of the problem to be covered by the Laboratory were possible ``fixes`` to the control rod sticking problem caused by VSR channel distortion.

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.

Status and problems of fusion reactor development.

PubMed

Schumacher, U

2001-03-01

Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

Lévitation magnétique par association d'aimants permanents et de supraconducteurs à haute température critique

NASA Astrophysics Data System (ADS)

Hiebel, P.; Tixador, P.; Chaud, X.

1995-06-01

Since their discovery in the years 1986/87, the high critical temperature superconductors have reached nowadays performances interesting enough to conceive passive magnetic bearings and suspensions which would combined permanent magnets and naturally stable superconducting pellets. After underlining the principal factors that affect the superconductormagnet interaction, different experimental results are given about vertical and axial forces with some stiffness values. The magnetization curve of a superconductor help to understand the hysteretic behavior of the force as a function of the distance between superconductor and magnet. So called simple and hybrid structures of superconducting magnetic suspension are presented. Finally simple numerical simulations allow to draw some interesting conclusions about both geometry and best fitting structure of permanent magnets. Depuis leur découverte dans les années 1986/87, les supraconducteurs à haute température critique ont désormais atteint des performances intéressantes et rendent envisageables des paliers et suspensions magnétiques passives associant aimants permanents et pastilles supraconductrices naturellement stables. Après avoir indiqué les termes importants influençant l'interaction supraconducteur - aimant, différents relevés expérimentaux sont donnés pour les forces verticales et transversales avec quelques valeurs de raideurs. La courbe d'aimantation d'un supraconducteur permet de comprendre le comportement hystérétique de la force en fonction de la distance supraconducteur-aimant. Les structures dites simple et hybride des suspensions magnétiques supraconductrices sont présentées. Enfin quelques simulations numériques simples permettent de dégager quelques conclusions intéressantes quant aux géométries respectives et aux structures d'aimants permanents les mieux adaptées.

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

Research reactor loading pattern optimization using estimation of distribution algorithms

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

Jiang, S.; Ziver, K.; AMCG Group, RM Consultants, Abingdon

2006-07-01

A new evolutionary search based approach for solving the nuclear reactor loading pattern optimization problems is presented based on the Estimation of Distribution Algorithms. The optimization technique developed is then applied to the maximization of the effective multiplication factor (K{sub eff}) of the Imperial College CONSORT research reactor (the last remaining civilian research reactor in the United Kingdom). A new elitism-guided searching strategy has been developed and applied to improve the local convergence together with some problem-dependent information based on the 'stand-alone K{sub eff} with fuel coupling calculations. A comparison study between the EDAs and a Genetic Algorithm with Heuristicmore » Tie Breaking Crossover operator has shown that the new algorithm is efficient and robust. (authors)« less

Shift Verification and Validation

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

Pandya, Tara M.; Evans, Thomas M.; Davidson, Gregory G

2016-09-07

This documentation outlines the verification and validation of Shift for the Consortium for Advanced Simulation of Light Water Reactors (CASL). Five main types of problems were used for validation: small criticality benchmark problems; full-core reactor benchmarks for light water reactors; fixed-source coupled neutron-photon dosimetry benchmarks; depletion/burnup benchmarks; and full-core reactor performance benchmarks. We compared Shift results to measured data and other simulated Monte Carlo radiation transport code results, and found very good agreement in a variety of comparison measures. These include prediction of critical eigenvalue, radial and axial pin power distributions, rod worth, leakage spectra, and nuclide inventories over amore » burn cycle. Based on this validation of Shift, we are confident in Shift to provide reference results for CASL benchmarking.« less

Discrete Ordinate Quadrature Selection for Reactor-based Eigenvalue Problems

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

Jarrell, Joshua J; Evans, Thomas M; Davidson, Gregory G

2013-01-01

In this paper we analyze the effect of various quadrature sets on the eigenvalues of several reactor-based problems, including a two-dimensional (2D) fuel pin, a 2D lattice of fuel pins, and a three-dimensional (3D) reactor core problem. While many quadrature sets have been applied to neutral particle discrete ordinate transport calculations, the Level Symmetric (LS) and the Gauss-Chebyshev product (GC) sets are the most widely used in production-level reactor simulations. Other quadrature sets, such as Quadruple Range (QR) sets, have been shown to be more accurate in shielding applications. In this paper, we compare the LS, GC, QR, and themore » recently developed linear-discontinuous finite element (LDFE) sets, as well as give a brief overview of other proposed quadrature sets. We show that, for a given number of angles, the QR sets are more accurate than the LS and GC in all types of reactor problems analyzed (2D and 3D). We also show that the LDFE sets are more accurate than the LS and GC sets for these problems. We conclude that, for problems where tens to hundreds of quadrature points (directions) per octant are appropriate, QR sets should regularly be used because they have similar integration properties as the LS and GC sets, have no noticeable impact on the speed of convergence of the solution when compared with other quadrature sets, and yield more accurate results. We note that, for very high-order scattering problems, the QR sets exactly integrate fewer angular flux moments over the unit sphere than the GC sets. The effects of those inexact integrations have yet to be analyzed. We also note that the LDFE sets only exactly integrate the zeroth and first angular flux moments. Pin power comparisons and analyses are not included in this paper and are left for future work.« less

Discrete ordinate quadrature selection for reactor-based Eigenvalue problems

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

Jarrell, J. J.; Evans, T. M.; Davidson, G. G.

2013-07-01

In this paper we analyze the effect of various quadrature sets on the eigenvalues of several reactor-based problems, including a two-dimensional (2D) fuel pin, a 2D lattice of fuel pins, and a three-dimensional (3D) reactor core problem. While many quadrature sets have been applied to neutral particle discrete ordinate transport calculations, the Level Symmetric (LS) and the Gauss-Chebyshev product (GC) sets are the most widely used in production-level reactor simulations. Other quadrature sets, such as Quadruple Range (QR) sets, have been shown to be more accurate in shielding applications. In this paper, we compare the LS, GC, QR, and themore » recently developed linear-discontinuous finite element (LDFE) sets, as well as give a brief overview of other proposed quadrature sets. We show that, for a given number of angles, the QR sets are more accurate than the LS and GC in all types of reactor problems analyzed (2D and 3D). We also show that the LDFE sets are more accurate than the LS and GC sets for these problems. We conclude that, for problems where tens to hundreds of quadrature points (directions) per octant are appropriate, QR sets should regularly be used because they have similar integration properties as the LS and GC sets, have no noticeable impact on the speed of convergence of the solution when compared with other quadrature sets, and yield more accurate results. We note that, for very high-order scattering problems, the QR sets exactly integrate fewer angular flux moments over the unit sphere than the GC sets. The effects of those inexact integrations have yet to be analyzed. We also note that the LDFE sets only exactly integrate the zeroth and first angular flux moments. Pin power comparisons and analyses are not included in this paper and are left for future work. (authors)« 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.

Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

NASA Astrophysics Data System (ADS)

Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

2016-06-01

Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

New Technological Platform for the National Nuclear Energy Strategy Development

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Rachkov, V. I.

2017-12-01

The paper considers the need to update the development strategy of Russia's nuclear power industry and various approaches to the large-scale nuclear power development. Problems of making decisions on fast neutron reactors and closed nuclear fuel cycle (NFC) arrangement are discussed. The current state of the development of fast neutron reactors and closed NFC technologies in Russia is considered and major problems are highlighted.

Comparing the new generation accelerator driven subcritical reactor system (ADS) to traditional critical reactors

NASA Astrophysics Data System (ADS)

Kemah, Elif; Akkaya, Recep; Tokgöz, Seyit Rıza

2017-02-01

In recent years, the accelerator driven subcritical reactors have taken great interest worldwide. The Accelerator Driven System (ADS) has been used to produce neutron in subcritical state by the external proton beam source. These reactors, which are hybrid systems, are important in production of clean and safe energy and conversion of radioactive waste. The ADS with the selection of reliability and robust target materials have been the new generation of fission reactors. In addition, in the ADS Reactors the problems of long-lived radioactive fission products and waste actinides encountered in the fission process of the reactor during incineration can be solved, and ADS has come to the forefront of thorium as fuel for the reactors.

Energy production using fission fragment rockets

NASA Astrophysics Data System (ADS)

Chapline, G.; Matsuda, Y.

1991-08-01

Fission fragment rockets are nuclear reactors with a core consisting of thin fibers in a vacuum, and which use magnetic fields to extract the fission fragments from the reactor core. As an alternative to ordinary nuclear reactors, fission fragment rockets would have the following advantages: approximately twice the efficiency if the fission fragment energy can be directly converted into electricity; reduction of the buildup of a fission fragment inventory in the reactor could avoid a Chernobyl type disaster; and collection of the fission fragments outside the reactor could simplify the waste disposal problem.

Experiment for search for sterile neutrino at SM-3 reactor

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

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.

Traduction et langues de specialite: Approches theoriques et considerations pedagogiques (Translation and Specialty Languages: Theoretical Approaches and Pedagogic Considerations).

ERIC Educational Resources Information Center

Guevel, Zelie, Ed.; Valentine, Egan, Ed.

Essays on the teaching of translation and on specialized translation, all in French, include: "Perspectives d'optimisation de la formation du traducteur: quelques reflexions" ("Perspectives on Optimization of Training of Translation Teachers: Some Reflections") (Egan Valentine); "L'enseignement de la revision…

Psychologie des discours et didactique des textes (Psychology of Discourse and the Teaching of Texts).

ERIC Educational Resources Information Center

Bronckart, Jean-Paul, Ed.

1995-01-01

This collection of articles on the nature of discourse and writing instruction include: "Une demarche de psychologie de discours; quelques aspects introductifs" ("An Application of Discourse Psychology; Introductory Thoughts") (Jean-Paul Bronckart); "Les procedes de prise en charge enonciative dans trois genres de texts expositifs" ("The Processes…

[Not Available].

PubMed

Ettalbi, S; Ibnouzahir, M; Droussi, H; Wahbi, S; Bahaichar, N; Boukind, E H

2009-06-30

La brûlure est un accident qui reste toujours très fréquent au Maroc, ce qui fait d'elle un problème de la santé publique. Les brûlures, quand elles sont graves ou profondes, entraînent de façon quasi inéluctable des séquelles fonctionnelles et esthétiques. A travers deux observations de deux enfants présentant des séquelles de brûlures graves, ayant retenti péjorativement sur leurs scolarités, on a essayé de mettre en évidence quelques facteurs incriminés dans cette tragédie (feu, petites bouteilles de gaz et le manque d'infrastructure, du personnel médical et paramédical, du matériel ainsi que de la prévention) comme étant une grande cause dans la survenue de ces séquelles. Le but de notre travail est d'énumérer ces différents facteurs intriqués, ainsi que de proposer quelques solutions, tout en insistant sur la prévention.

Reactor operations informal monthly report, May 1, 1995--May 31, 1995

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

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-05-01

This document is an informal progress report for the operational performance of the Brookhaven Medical Research Reactor, and the Brookhaven High Flux Beam Reactor, for the month of May, 1995. Both machines ran well during this period, with no reportable instrumentation problems, all scheduled maintenance performed, and only one reportable occurance, involving a particle on Vest Button, Personnel Radioactive Contamination.

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

NASA Astrophysics Data System (ADS)

Drajat, R. Z.; Su'ud, Z.; Soewono, E.; Gunawan, A. Y.

2012-05-01

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

Performance Assessment of the Commercial CFD Software for the Prediction of the Reactor Internal Flow

NASA Astrophysics Data System (ADS)

Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong; Kim, Do Hyeong; Kang, Min Ku

2014-06-01

As the computer hardware technology develops the license applicants for nuclear power plant use the commercial CFD software with the aim of reducing the excessive conservatism associated with using simplified and conservative analysis tools. Even if some of CFD software developer and its user think that a state of the art CFD software can be used to solve reasonably at least the single-phase nuclear reactor problems, there is still limitation and uncertainty in the calculation result. From a regulatory perspective, Korea Institute of Nuclear Safety (KINS) is presently conducting the performance assessment of the commercial CFD software for nuclear reactor problems. In this study, in order to examine the validity of the results of 1/5 scaled APR+ (Advanced Power Reactor Plus) flow distribution tests and the applicability of CFD in the analysis of reactor internal flow, the simulation was conducted with the two commercial CFD software (ANSYS CFX V.14 and FLUENT V.14) among the numerous commercial CFD software and was compared with the measurement. In addition, what needs to be improved in CFD for the accurate simulation of reactor core inlet flow was discussed.

Self-teaching neural network learns difficult reactor control problem

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

Jouse, W.C.

1989-01-01

A self-teaching neural network used as an adaptive controller quickly learns to control an unstable reactor configuration. The network models the behavior of a human operator. It is trained by allowing it to operate the reactivity control impulsively. It is punished whenever either the power or fuel temperature stray outside technical limits. Using a simple paradigm, the network constructs an internal representation of the punishment and of the reactor system. The reactor is constrained to small power orbits.

Status report on the disposal of radioactive wastes

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

Culler, F.L. Jr.; McLain, S.

1957-06-25

A comprehensive survey of waste disposal techniques, requirements, costs, hazards, and long-range considerations is presented. The nature of high level wastes from reactors and chemical processes, in the form of fission product gases, waste solutions, solid wastes, and particulate solids in gas phase, is described. Growth predictions for nuclear reactor capacity and the associated fission product and transplutonic waste problem are made and discussed on the basis of present knowledge. Biological hazards from accumulated wastes and potential hazards from reactor accidents, ore and feed material processing, chemical reprocessing plants, and handling of fissionable and fertile material after irradiation and decontaminationmore » are surveyed. The waste transportation problem is considered from the standpoints of magnitude of the problem, present regulations, costs, and cooling periods. The possibilities for ultimate waste management and/or disposal are reviewed and discussed. The costs of disposal, evaporation, storage tanks, and drum-drying are considered.« less

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

L'acquisition d'une language seconde: Quelques developpements theoriques recents (Second Language Acquisition: Some Recent Theoretical Developments).

ERIC Educational Resources Information Center

Py, Bernard, Ed.

1994-01-01

This collection of articles on second language learning includes: "Action, langage et discours. Les fondements d'une psychologie du langage" ("Action, Language, and Discourse. Foundations of a Psychology of Language") (Jean-Paul Bronckart); "Contextes socio-culturels et appropriation des languages secondes: l'apprentissage en milieu social et la…

Canadian Association for the Study of Adult Education. Proceedings of the Annual Conference (4th, Montreal, Quebec, Canada, May 28-30, 1985).

ERIC Educational Resources Information Center

Canadian Association for the Study of Adult Education, Guelph (Ontario).

These proceedings contain 28 papers (20 in English and 8 in French), including the following: "Beyond Ideology: The Case of the Corporate Classroom" (Zinman); "De quelques dimensions paradoxales de l'education interculturelle" (Ollivier); "Ideology, Indoctrination and the Language of Physics" (Winchester);…

Radiation effect of neutrons produced by D-D side reactions on a D-3He fusion reactor

NASA Astrophysics Data System (ADS)

Bahmani, J.

2017-04-01

One of the most important characteristics in D-3He fusion reactors is neutron production via D-D side reactions. The neutrons can activate structural material, degrading them and ultimately converting them into high-level radioactive waste, while it is really costly and difficult to remove them. The neutrons from a fusion reactor could also be used to make weapons-grade nuclear material, rendering such types of fusion reactors a serious proliferation hazard. A related problem is the presence of radioactive elements such as tritium in D-3He plasma, either as fuel for or as products of the nuclear reactions; substantial quantities of radioactive elements would not only pose a general health risk, but tritium in particular would also be another proliferation hazard. The problems of neutron radiation and radioactive element production are especially interconnected because both would result from the D-D side reaction. Therefore, the presentation approach for reducing neutrons via D-D nuclear side reactions in a D-3He fusion reactor is very important. For doing this research, energy losses and neutron power fraction in D-3He fusion reactors are investigated. Calculations show neutrons produced by the D-D nuclear side reaction could be reduced by changing to a more 3He-rich fuel mixture, but then the bremsstrahlung power loss fraction would increase in the D-3He fusion reactor.

Multivariable optimization of an auto-thermal ammonia synthesis reactor using genetic algorithm

NASA Astrophysics Data System (ADS)

Anh-Nga, Nguyen T.; Tuan-Anh, Nguyen; Tien-Dung, Vu; Kim-Trung, Nguyen

2017-09-01

The ammonia synthesis system is an important chemical process used in the manufacture of fertilizers, chemicals, explosives, fibers, plastics, refrigeration. In the literature, many works approaching the modeling, simulation and optimization of an auto-thermal ammonia synthesis reactor can be found. However, they just focus on the optimization of the reactor length while keeping the others parameters constant. In this study, the other parameters are also considered in the optimization problem such as the temperature of feed gas enters the catalyst zone. The optimal problem requires the maximization of a multivariable objective function which subjects to a number of equality constraints involving the solution of coupled differential equations and also inequality constraints. The solution of an optimization problem can be found through, among others, deterministic or stochastic approaches. The stochastic methods, such as evolutionary algorithm (EA), which is based on natural phenomenon, can overcome the drawbacks such as the requirement of the derivatives of the objective function and/or constraints, or being not efficient in non-differentiable or discontinuous problems. Genetic algorithm (GA) which is a class of EA, exceptionally simple, robust at numerical optimization and is more likely to find a true global optimum. In this study, the genetic algorithm is employed to find the optimum profit of the process. The inequality constraints were treated using penalty method. The coupled differential equations system was solved using Runge-Kutta 4th order method. The results showed that the presented numerical method could be applied to model the ammonia synthesis reactor. The optimum economic profit obtained from this study are also compared to the results from the literature. It suggests that the process should be operated at higher temperature of feed gas in catalyst zone and the reactor length is slightly longer.

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

PubMed

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

2017-08-26

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

Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

NASA Astrophysics Data System (ADS)

Shchelik, S. V.; Pavlov, A. S.

2013-07-01

Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

Reactivity change in a fast-spectrum space power reactor due to a 328-meter-per-second (1075-ft/sec) impact

NASA Technical Reports Server (NTRS)

Peoples, J. A., Jr.; Puthoff, R. L.

1973-01-01

Application of nuclear reactors in space will present operational problems. One such problem is the possibility of an earth impact at velocities in excess of 305 m/sec (1000 ft/sec). This report shows the results of an impact against concrete at 328 m/sec (1075 ft/sec) and examines the deformed core to estimate the range of activity inserted as a result of the impact. The results of this examination are that the deformation of the reactor core within the containment vessel left only an estimated 2.7 percent void in the core and that the reactivity inserted due to this impact deformation could be from 4.0 to 10.25 dollars.

On Heat Loading, Novel Divertors, and Fusion Reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike

2006-10-01

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction (fRad) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of fRad will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core fRad within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.

An approach for coupled-code multiphysics core simulations from a common input

DOE PAGES

Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; ...

2014-12-10

This study describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which ismore » built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Finally, ongoing development of this approach is also briefly described.« less

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

NASA Astrophysics Data System (ADS)

Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-01

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

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

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclearmore » waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.« less

Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

PubMed

Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

2018-01-04

A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

A propos de quelques experiences de francais fonctionnel en milieu hispanaphone (A Commentary on Experiments in Functional French in a Spanish-Speaking Milieu)

ERIC Educational Resources Information Center

Baltzer, Francois

1978-01-01

A discussion of the adaptation of audiovisual methods to respond to various specific needs in Mexico City. Some of the topics discussed are: meeting needs of people involved in special fields, particularly science, technology and economics; and the use of television for functional French instruction. (AMH)

Aspect Epidemiologique des Sequelles de Brulures a Marrakech, Maroc, a Travers Deux Observations

PubMed Central

Ettalbi, S.; Ibnouzahir, M.; Droussi, H.; Wahbi, S.; Bahaichar, N.; Boukind, E.H.

2009-01-01

Summary La brûlure est un accident qui reste toujours très fréquent au Maroc, ce qui fait d'elle un problème de la santé publique. Les brûlures, quand elles sont graves ou profondes, entraînent de façon quasi inéluctable des séquelles fonctionnelles et esthétiques. A travers deux observations de deux enfants présentant des séquelles de brûlures graves, ayant retenti péjorativement sur leurs scolarités, on a essayé de mettre en évidence quelques facteurs incriminés dans cette tragédie (feu, petites bouteilles de gaz et le manque d'infrastructure, du personnel médical et paramédical, du matériel ainsi que de la prévention) comme étant une grande cause dans la survenue de ces séquelles. Le but de notre travail est d'énumérer ces différents facteurs intriqués, ainsi que de proposer quelques solutions, tout en insistant sur la prévention. PMID:21991156

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

Effect of shear stress on cell cultures and other reactor problems

NASA Technical Reports Server (NTRS)

Schleier, H.

1981-01-01

Anchorage dependent cell cultures in fluidized beds are tested. Feasibility calculations indicate the allowed parameters and estimate the shear stresses therein. In addition, the diffusion equation with first order reaction is solved for the spherical shell (double bubble) reactor with various constraints.

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

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

Velikhov, E. P.; Kovalchuk, M. V.; Azizov, E. A., E-mail: Azizov-EA@nrcki.ru

2015-12-15

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel canmore » be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.« less

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

Design of a tokamak fusion reactor first wall armor against neutral beam impingement

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

Myers, R.A.

1977-12-01

The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiationmore » damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem.« less

Reactor Application for Coaching Newbies

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

2015-06-17

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

Application of CFX-10 to the Investigation of RPV Coolant Mixing in VVER Reactors

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

Moretti, Fabio; Melideo, Daniele; Terzuoli, Fulvio

2006-07-01

Coolant mixing phenomena occurring in the pressure vessel of a nuclear reactor constitute one of the main objectives of investigation by researchers concerned with nuclear reactor safety. For instance, mixing plays a relevant role in reactivity-induced accidents initiated by de-boration or boron dilution events, followed by transport of a de-borated slug into the vessel of a pressurized water reactor. Another example is constituted by temperature mixing, which may sensitively affect the consequences of a pressurized thermal shock scenario. Predictive analysis of mixing phenomena is strongly improved by the availability of computational tools able to cope with the inherent three-dimensionality ofmore » such problem, like system codes with three-dimensional capabilities, and Computational Fluid Dynamics (CFD) codes. The present paper deals with numerical analyses of coolant mixing in the reactor pressure vessel of a VVER-1000 reactor, performed by the ANSYS CFX-10 CFD code. In particular, the 'swirl' effect that has been observed to take place in the downcomer of such kind of reactor has been addressed, with the aim of assessing the capability of the codes to predict that effect, and to understand the reasons for its occurrence. Results have been compared against experimental data from V1000CT-2 Benchmark. Moreover, a boron mixing problem has been investigated, in the hypothesis that a de-borated slug, transported by natural circulation, enters the vessel. Sensitivity analyses have been conducted on some geometrical features, model parameters and boundary conditions. (authors)« less

MYRRHA: A multipurpose nuclear research facility

NASA Astrophysics Data System (ADS)

Baeten, P.; Schyns, M.; Fernandez, Rafaël; De Bruyn, Didier; Van den Eynde, Gert

2014-12-01

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a multipurpose research facility currently being developed at SCK•CEN. MYRRHA is based on the ADS (Accelerator Driven System) concept where a proton accelerator, a spallation target and a subcritical reactor are coupled. MYRRHA will demonstrate the ADS full concept by coupling these three components at a reasonable power level to allow operation feedback. As a flexible irradiation facility, the MYRRHA research facility will be able to work in both critical as subcritical modes. In this way, MYRRHA will allow fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. MYRRHA will be cooled by lead-bismuth eutectic and will play an important role in the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) GEN IV concept. MYRRHA will also contribute to the study of partitioning and transmutation of high-level waste. Transmutation of minor actinides (MA) can be completed in an efficient way in fast neutron spectrum facilities, so both critical reactors and subcritical ADS are potential candidates as dedicated transmutation systems. However critical reactors heavily loaded with fuel containing large amounts of MA pose reactivity control problems, and thus safety problems. A subcritical ADS operates in a flexible and safe manner, even with a core loading containing a high amount of MA leading to a high transmutation rate. In this paper, the most recent developments in the design of the MYRRHA facility are presented.

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

Craig, D.F.

The division was formed in 1946 at the suggestion of Dr. Eugene P. Wigner to attack the problem of the distortion of graphite in the early reactors due to exposure to reactor neutrons, and the consequent radiation damage. It was called the Metallurgy Division and assembled the metallurgical and solid state physics activities of the time which were not directly related to nuclear weapons production. William A. Johnson, a Westinghouse employee, was named Division Director in 1946. In 1949 he was replaced by John H Frye Jr. when the Division consisted of 45 people. He was director during most ofmore » what is called the Reactor Project Years until 1973 and his retirement. During this period the Division evolved into three organizational areas: basic research, applied research in nuclear reactor materials, and reactor programs directly related to a specific reactor(s) being designed or built. The Division (Metals and Ceramics) consisted of 204 staff members in 1973 when James R. Weir, Jr., became Director. This was the period of the oil embargo, the formation of the Energy Research and Development Administration (ERDA) by combining the Atomic Energy Commission (AEC) with the Office of Coal Research, and subsequent formation of the Department of Energy (DOE). The diversification process continued when James O. Stiegler became Director in 1984, partially as a result of the pressure of legislation encouraging the national laboratories to work with U.S. industries on their problems. During that time the Division staff grew from 265 to 330. Douglas F. Craig became Director in 1992.« less

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

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

Díez, C.J., E-mail: cj.diez@upm.es; Cabellos, O.; Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has tomore » be performed in order to analyse the limitations of using one-group uncertainties.« less

Beryllium for fusion application - recent results

NASA Astrophysics Data System (ADS)

Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

2002-12-01

The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described.

Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

NASA Astrophysics Data System (ADS)

Díez, C. J.; Cabellos, O.; Martínez, J. S.

2015-01-01

Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

Radiocarbon tracer measurements of atmospheric hydroxyl radical concentrations

NASA Technical Reports Server (NTRS)

Campbell, M. J.; Farmer, J. C.; Fitzner, C. A.; Henry, M. N.; Sheppard, J. C.

1986-01-01

The usefulness of the C-14 tracer in measurements of atmospheric hydroxyl radical concentration is discussed. The apparatus and the experimental conditions of three variations of a radiochemical method of atmosphere analysis are described and analyzed: the Teflon bag static reactor, the flow reactor (used in the Wallops Island tests), and the aircraft OH titration reactor. The procedure for reduction of the aircraft reactor instrument data is outlined. The problems connected with the measurement of hydroxyl radicals are discussed. It is suggested that the gas-phase radioisotope methods have considerable potential in measuring tropospheric impurities present in very low concentrations.

Developments and Tendencies in Fission Reactor Concepts

NASA Astrophysics Data System (ADS)

Adamov, E. O.; Fuji-Ie, Y.

This chapter describes, in two parts, new-generation nuclear energy systems that are required to be in harmony with nature and to make full use of nuclear resources. The issues of transmutation and containment of radioactive waste will also be addressed. After a short introduction to the first part, Sect. 58.1.2 will detail the requirements these systems must satisfy on the basic premise of peaceful use of nuclear energy. The expected designs themselves are described in Sect. 58.1.3. The subsequent sections discuss various types of advanced reactor systems. Section 58.1.4 deals with the light water reactor (LWR) whose performance is still expected to improve, which would extend its application in the future. The supercritical-water-cooled reactor (SCWR) will also be shortly discussed. Section 58.1.5 is mainly on the high temperature gas-cooled reactor (HTGR), which offers efficient and multipurpose use of nuclear energy. The gas-cooled fast reactor (GFR) is also included. Section 58.1.6 focuses on the sodium-cooled fast reactor (SFR) as a promising concept for advanced nuclear reactors, which may help both to achieve expansion of energy sources and environmental protection thus contributing to the sustainable development of mankind. The molten-salt reactor (MSR) is shortly described in Sect. 58.1.7. The second part of the chapter deals with reactor systems of a new generation, which are now found at the research and development (R&D) stage and in the medium term of 20-30 years can shape up as reliable, economically efficient, and environmentally friendly energy sources. They are viewed as technologies of cardinal importance, capable of resolving the problems of fuel resources, minimizing the quantities of generated radioactive waste and the environmental impacts, and strengthening the regime of nonproliferation of the materials suitable for nuclear weapons production. Particular attention has been given to naturally safe fast reactors with a closed fuel cycle (CFC) - as an advanced and promising reactor system that offers solutions to the above problems. The difference (not confrontation) between the approaches to nuclear power development based on the principles of “inherent safety” and “natural safety” is demonstrated.

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).

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

PubMed

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

2018-08-01

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

Support vector machines for nuclear reactor state estimation

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

Zavaljevski, N.; Gross, K. C.

2000-02-14

Validation of nuclear power reactor signals is often performed by comparing signal prototypes with the actual reactor signals. The signal prototypes are often computed based on empirical data. The implementation of an estimation algorithm which can make predictions on limited data is an important issue. A new machine learning algorithm called support vector machines (SVMS) recently developed by Vladimir Vapnik and his coworkers enables a high level of generalization with finite high-dimensional data. The improved generalization in comparison with standard methods like neural networks is due mainly to the following characteristics of the method. The input data space is transformedmore » into a high-dimensional feature space using a kernel function, and the learning problem is formulated as a convex quadratic programming problem with a unique solution. In this paper the authors have applied the SVM method for data-based state estimation in nuclear power reactors. In particular, they implemented and tested kernels developed at Argonne National Laboratory for the Multivariate State Estimation Technique (MSET), a nonlinear, nonparametric estimation technique with a wide range of applications in nuclear reactors. The methodology has been applied to three data sets from experimental and commercial nuclear power reactor applications. The results are promising. The combination of MSET kernels with the SVM method has better noise reduction and generalization properties than the standard MSET algorithm.« less

Root-cause analysis of the better performance of the coarse-mesh finite-difference method for CANDU-type reactors

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

Shen, W.

2012-07-01

Recent assessment results indicate that the coarse-mesh finite-difference method (FDM) gives consistently smaller percent differences in channel powers than the fine-mesh FDM when compared to the reference MCNP solution for CANDU-type reactors. However, there is an impression that the fine-mesh FDM should always give more accurate results than the coarse-mesh FDM in theory. To answer the question if the better performance of the coarse-mesh FDM for CANDU-type reactors was just a coincidence (cancellation of errors) or caused by the use of heavy water or the use of lattice-homogenized cross sections for the cluster fuel geometry in the diffusion calculation, threemore » benchmark problems were set up with three different fuel lattices: CANDU, HWR and PWR. These benchmark problems were then used to analyze the root cause of the better performance of the coarse-mesh FDM for CANDU-type reactors. The analyses confirm that the better performance of the coarse-mesh FDM for CANDU-type reactors is mainly caused by the use of lattice-homogenized cross sections for the sub-meshes of the cluster fuel geometry in the diffusion calculation. Based on the analyses, it is recommended to use 2 x 2 coarse-mesh FDM to analyze CANDU-type reactors when lattice-homogenized cross sections are used in the core analysis. (authors)« less

Quelques Facteurs Sociaux Agissant sur la Formation Permanente et l'Education Informelle en Algerie (Social Factors Acting upon Lifelong Learning and Informal Education in Algeria).

ERIC Educational Resources Information Center

Haddab, Mustapha

1994-01-01

Analyzes conditions that have led to an increase in private and collective educational initiatives in Algeria, highlighting political and socioeconomic changes since 1988. Indicates that after a long period of a public education monopoly, social factors have led to the development of alternative educational opportunities that are more responsive…

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

Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

Olson, D.L.

1992-01-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory's Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

Surveillance application using patten recognition software at the EBR-II Reactor Facility

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

Olson, D.L.

1992-05-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory`s Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodiummore » Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller.« less

Heat transfer evaluation in a plasma core reactor

NASA Technical Reports Server (NTRS)

Smith, D. E.; Smith, T. M.; Stoenescu, M. L.

1976-01-01

Numerical evaluations of heat transfer in a fissioning uranium plasma core reactor cavity, operating with seeded hydrogen propellant, was performed. A two-dimensional analysis is based on an assumed flow pattern and cavity wall heat exchange rate. Various iterative schemes were required by the nature of the radiative field and by the solid seed vaporization. Approximate formulations of the radiative heat flux are generally used, due to the complexity of the solution of a rigorously formulated problem. The present work analyzes the sensitivity of the results with respect to approximations of the radiative field, geometry, seed vaporization coefficients and flow pattern. The results present temperature, heat flux, density and optical depth distributions in the reactor cavity, acceptable simplifying assumptions, and iterative schemes. The present calculations, performed in cartesian and spherical coordinates, are applicable to any most general heat transfer problem.

61. VIEW LOOKING NORTHWEST AT A SIGNAL REACTOR OR CHOKE ...

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

61. VIEW LOOKING NORTHWEST AT A SIGNAL REACTOR OR CHOKE COIL. WITHIN THE PROTECTIVE ENCLOSURE IS AN AIR AND PORCELAIN INSULATED COIL OF 5/8' DIAMETER STRANDED COPPER WIRE. REACTOR COILS WERE PLACED IN SERIES WITH EACH LEG OF THREE PHASE GENERATORS. THEIR FUNCTION WAS TO MODERATE SURGES OF CURRENT CAUSED BY LIGHTNING STRIKES, OPEN OR SHORT CIRCUIT PROBLEMS ON THE LINE. - New York, New Haven & Hartford Railroad, Cos Cob Power Plant, Sound Shore Drive, Greenwich, Fairfield County, CT

Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

NASA Technical Reports Server (NTRS)

Weinstein, H.; Lavan, Z.

1975-01-01

Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

Eigenvalue Solvers for Modeling Nuclear Reactors on Leadership Class Machines

DOE PAGES

Slaybaugh, R. N.; Ramirez-Zweiger, M.; Pandya, Tara; ...

2018-02-20

In this paper, three complementary methods have been implemented in the code Denovo that accelerate neutral particle transport calculations with methods that use leadership-class computers fully and effectively: a multigroup block (MG) Krylov solver, a Rayleigh quotient iteration (RQI) eigenvalue solver, and a multigrid in energy (MGE) preconditioner. The MG Krylov solver converges more quickly than Gauss Seidel and enables energy decomposition such that Denovo can scale to hundreds of thousands of cores. RQI should converge in fewer iterations than power iteration (PI) for large and challenging problems. RQI creates shifted systems that would not be tractable without the MGmore » Krylov solver. It also creates ill-conditioned matrices. The MGE preconditioner reduces iteration count significantly when used with RQI and takes advantage of the new energy decomposition such that it can scale efficiently. Each individual method has been described before, but this is the first time they have been demonstrated to work together effectively. The combination of solvers enables the RQI eigenvalue solver to work better than the other available solvers for large reactors problems on leadership-class machines. Using these methods together, RQI converged in fewer iterations and in less time than PI for a full pressurized water reactor core. These solvers also performed better than an Arnoldi eigenvalue solver for a reactor benchmark problem when energy decomposition is needed. The MG Krylov, MGE preconditioner, and RQI solver combination also scales well in energy. Finally, this solver set is a strong choice for very large and challenging problems.« less

Eigenvalue Solvers for Modeling Nuclear Reactors on Leadership Class Machines

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

Slaybaugh, R. N.; Ramirez-Zweiger, M.; Pandya, Tara

In this paper, three complementary methods have been implemented in the code Denovo that accelerate neutral particle transport calculations with methods that use leadership-class computers fully and effectively: a multigroup block (MG) Krylov solver, a Rayleigh quotient iteration (RQI) eigenvalue solver, and a multigrid in energy (MGE) preconditioner. The MG Krylov solver converges more quickly than Gauss Seidel and enables energy decomposition such that Denovo can scale to hundreds of thousands of cores. RQI should converge in fewer iterations than power iteration (PI) for large and challenging problems. RQI creates shifted systems that would not be tractable without the MGmore » Krylov solver. It also creates ill-conditioned matrices. The MGE preconditioner reduces iteration count significantly when used with RQI and takes advantage of the new energy decomposition such that it can scale efficiently. Each individual method has been described before, but this is the first time they have been demonstrated to work together effectively. The combination of solvers enables the RQI eigenvalue solver to work better than the other available solvers for large reactors problems on leadership-class machines. Using these methods together, RQI converged in fewer iterations and in less time than PI for a full pressurized water reactor core. These solvers also performed better than an Arnoldi eigenvalue solver for a reactor benchmark problem when energy decomposition is needed. The MG Krylov, MGE preconditioner, and RQI solver combination also scales well in energy. Finally, this solver set is a strong choice for very large and challenging problems.« less

The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

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

Smith, R.R.

1986-01-01

This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

Design of a self-tuning regulator for temperature control of a polymerization reactor.

PubMed

Vasanthi, D; Pranavamoorthy, B; Pappa, N

2012-01-01

The temperature control of a polymerization reactor described by Chylla and Haase, a control engineering benchmark problem, is used to illustrate the potential of adaptive control design by employing a self-tuning regulator concept. In the benchmark scenario, the operation of the reactor must be guaranteed under various disturbing influences, e.g., changing ambient temperatures or impurity of the monomer. The conventional cascade control provides a robust operation, but often lacks in control performance concerning the required strict temperature tolerances. The self-tuning control concept presented in this contribution solves the problem. This design calculates a trajectory for the cooling jacket temperature in order to follow a predefined trajectory of the reactor temperature. The reaction heat and the heat transfer coefficient in the energy balance are estimated online by using an unscented Kalman filter (UKF). Two simple physically motivated relations are employed, which allow the non-delayed estimation of both quantities. Simulation results under model uncertainties show the effectiveness of the self-tuning control concept. Copyright © 2011 ISA. Published by Elsevier Ltd. All rights reserved.

Numerical Tests for the Problem of U-Pu Fuel Burnup in Fuel Rod and Polycell Models Using the MCNP Code

NASA Astrophysics Data System (ADS)

Muratov, V. G.; Lopatkin, A. V.

An important aspect in the verification of the engineering techniques used in the safety analysis of MOX-fuelled reactors, is the preparation of test calculations to determine nuclide composition variations under irradiation and analysis of burnup problem errors resulting from various factors, such as, for instance, the effect of nuclear data uncertainties on nuclide concentration calculations. So far, no universally recognized tests have been devised. A calculation technique has been developed for solving the problem using the up-to-date calculation tools and the latest versions of nuclear libraries. Initially, in 1997, a code was drawn up in an effort under ISTC Project No. 116 to calculate the burnup in one VVER-1000 fuel rod, using the MCNP Code. Later on, the authors developed a computation technique which allows calculating fuel burnup in models of a fuel rod, or a fuel assembly, or the whole reactor. It became possible to apply it to fuel burnup in all types of nuclear reactors and subcritical blankets.

Optimization of an auto-thermal ammonia synthesis reactor using cyclic coordinate method

NASA Astrophysics Data System (ADS)

A-N Nguyen, T.; Nguyen, T.-A.; Vu, T.-D.; Nguyen, K.-T.; K-T Dao, T.; P-H Huynh, K.

2017-06-01

The ammonia synthesis system is an important chemical process used in the manufacture of fertilizers, chemicals, explosives, fibers, plastics, refrigeration. In the literature, many works approaching the modeling, simulation and optimization of an auto-thermal ammonia synthesis reactor can be found. However, they just focus on the optimization of the reactor length while keeping the others parameters constant. In this study, the other parameters are also considered in the optimization problem such as the temperature of feed gas enters the catalyst zone, the initial nitrogen proportion. The optimal problem requires the maximization of an objective function which is multivariable function and subject to a number of equality constraints involving the solution of coupled differential equations and also inequality constraint. The cyclic coordinate search was applied to solve the multivariable-optimization problem. In each coordinate, the golden section method was applied to find the maximum value. The inequality constraints were treated using penalty method. The coupled differential equations system was solved using Runge-Kutta 4th order method. The results obtained from this study are also compared to the results from the literature.

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

NASA Astrophysics Data System (ADS)

Tabor, R. W.

1986-09-01

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

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

NASA Astrophysics Data System (ADS)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

Method of producing pyrolysis gases from carbon-containing materials

DOEpatents

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

1989-01-01

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

Nuclear reactor transient analysis via a quasi-static kinetics Monte Carlo method

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

Jo, YuGwon; Cho, Bumhee; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr

2015-12-31

The predictor-corrector quasi-static (PCQS) method is applied to the Monte Carlo (MC) calculation for reactor transient analysis. To solve the transient fixed-source problem of the PCQS method, fission source iteration is used and a linear approximation of fission source distributions during a macro-time step is introduced to provide delayed neutron source. The conventional particle-tracking procedure is modified to solve the transient fixed-source problem via MC calculation. The PCQS method with MC calculation is compared with the direct time-dependent method of characteristics (MOC) on a TWIGL two-group problem for verification of the computer code. Then, the results on a continuous-energy problemmore » are presented.« less

New core-reflector boundary conditions for transient nodal reactor calculations

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

Lee, E.K.; Kim, C.H.; Joo, H.K.

1995-09-01

New core-reflector boundary conditions designed for the exclusion of the reflector region in transient nodal reactor calculations are formulated. Spatially flat frequency approximations for the temporal neutron behavior and two types of transverse leakage approximations in the reflector region are introduced to solve the transverse-integrated time-dependent one-dimensional diffusion equation and then to obtain relationships between net current and flux at the core-reflector interfaces. To examine the effectiveness of new core-reflector boundary conditions in transient nodal reactor computations, nodal expansion method (NEM) computations with and without explicit representation of the reflector are performed for Laboratorium fuer Reaktorregelung und Anlagen (LRA) boilingmore » water reactor (BWR) and Nuclear Energy Agency Committee on Reactor Physics (NEACRP) pressurized water reactor (PWR) rod ejection kinetics benchmark problems. Good agreement between two NEM computations is demonstrated in all the important transient parameters of two benchmark problems. A significant amount of CPU time saving is also demonstrated with the boundary condition model with transverse leakage (BCMTL) approximations in the reflector region. In the three-dimensional LRA BWR, the BCMTL and the explicit reflector model computations differ by {approximately}4% in transient peak power density while the BCMTL results in >40% of CPU time saving by excluding both the axial and the radial reflector regions from explicit computational nodes. In the NEACRP PWR problem, which includes six different transient cases, the largest difference is 24.4% in the transient maximum power in the one-node-per-assembly B1 transient results. This difference in the transient maximum power of the B1 case is shown to reduce to 11.7% in the four-node-per-assembly computations. As for the computing time, BCMTL is shown to reduce the CPU time >20% in all six transient cases of the NEACRP PWR.« less

Experiment neutrino-4 on searching for a sterile neutrino with multisection detector model

NASA Astrophysics Data System (ADS)

Serebrov, A. P.; Ivochkin, V. G.; Samoilov, R. M.; Fomin, A. K.; Zinov'ev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Chernyi, A. V.; Zherebtsov, O. M.; Polyushkin, A. O.; Martem'yanov, V. P.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Izhutov, A. L.; Tuzov, A. A.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanas'ev, V. V.; Zaitsev, M. E.; Chaikovskii, M. E.

2017-02-01

A laboratory for searching for oscillations of reactor antineutrinos has been created based on the SM-3 reactor in order to approach the problem of the possible existence of a sterile neutrino. The multisection detector prototype with a liquid scintillator volume of 350 L was installed in mid-2015. This detector can move inside the passive shield in a range of 6-11 m from the active core of the reactor. The antineutrino flux was measured for the first time at these short distances from the active core of the reactor by the movable detector. The measurements with the multisection detector prototype demonstrated that it is possible to measure the antineutrino flux from the reactor in the complicated conditions of cosmic background on the Earth's surface.

Support de la famille dans l'education: Quelques aspects de la realite grecque (Support of the Family for Education: Aspects of the Situation in Greece).

ERIC Educational Resources Information Center

Varnava-Skoura, Gella

1992-01-01

Describes extended family structure in Greece and offers a profile of the family backgrounds of university students. Finds that the cultural capital and sociolinguistic codes of families are not determining factors for university entry in Greece. University students come from clerical and mixed families, who are willing to make necessary financial…

Quelques applications de la decomposition de l'entropie en psychologie et pedagogie (Some Applications of the Resolution of Entropy in Psychology and Pedagogy)

ERIC Educational Resources Information Center

Gillet, Louis

1971-01-01

Psychological and educational measurement is carried out according to the type of model used and data collected. The H entropy which shows the dispersion of the data can be divided into intragroup and intergroup entropy. Choice of colors, sociometrical choice, and the communications are three situations where this resolution can be applied. (MF)

Le grand séisme de Huaxian (1556) : quelques documents chinois

NASA Astrophysics Data System (ADS)

Poirier, Jean-Paul

2017-03-01

The strong earthquake that struck Shaanxi, Shanxi and several other Chinese provinces in 1556 is generally considered as the deadliest of all earthquakes. It is said that the Chinese annals reported 830,000 casualties. We give here a translation into French of the relevant passage of the annals, as well as of a testimony of a survivor Qin Keda, and of a text engraved on a stela.

Quelques aspects du folklore de la region Roannaise autour de 1950 (Some Aspects of Folklore of the Roanne Region about 1950).

ERIC Educational Resources Information Center

Long, Jacqueline

1971-01-01

This article examines several aspects of folklore characteristic of the region of Roanne, France, during the 1950's. The town of Roanne, located between Clermont Ferrand and Lyon on the Loire River, is described in terms of its festive activities during serveral key holidays. The erosion of various customs and traditions, an inevitable result of…

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

NONE

Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of researchmore » reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)« less

A systematic reactor design approach for the synthesis of active pharmaceutical ingredients.

PubMed

Emenike, Victor N; Schenkendorf, René; Krewer, Ulrike

2018-05-01

Today's highly competitive pharmaceutical industry is in dire need of an accelerated transition from the drug development phase to the drug production phase. At the heart of this transition are chemical reactors that facilitate the synthesis of active pharmaceutical ingredients (APIs) and whose design can affect subsequent processing steps. Inspired by this challenge, we present a model-based approach for systematic reactor design. The proposed concept is based on the elementary process functions (EPF) methodology to select an optimal reactor configuration from existing state-of-the-art reactor types or can possibly lead to the design of novel reactors. As a conceptual study, this work summarizes the essential steps in adapting the EPF approach to optimal reactor design problems in the field of API syntheses. Practically, the nucleophilic aromatic substitution of 2,4-difluoronitrobenzene was analyzed as a case study of pharmaceutical relevance. Here, a small-scale tubular coil reactor with controlled heating was identified as the optimal set-up reducing the residence time by 33% in comparison to literature values. Copyright © 2017 Elsevier B.V. All rights reserved.

BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

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

Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

1981-06-01

This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

Optimal Refueling Pattern Search for a CANDU Reactor Using a Genetic Algorithm

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

Quang Binh, DO; Gyuhong, ROH; Hangbok, CHOI

2006-07-01

This paper presents the results from the application of genetic algorithms to a refueling optimization of a Canada deuterium uranium (CANDU) reactor. This work aims at making a mathematical model of the refueling optimization problem including the objective function and constraints and developing a method based on genetic algorithms to solve the problem. The model of the optimization problem and the proposed method comply with the key features of the refueling strategy of the CANDU reactor which adopts an on-power refueling operation. In this study, a genetic algorithm combined with an elitism strategy was used to automatically search for themore » refueling patterns. The objective of the optimization was to maximize the discharge burn-up of the refueling bundles, minimize the maximum channel power, or minimize the maximum change in the zone controller unit (ZCU) water levels. A combination of these objectives was also investigated. The constraints include the discharge burn-up, maximum channel power, maximum bundle power, channel power peaking factor and the ZCU water level. A refueling pattern that represents the refueling rate and channels was coded by a one-dimensional binary chromosome, which is a string of binary numbers 0 and 1. A computer program was developed in FORTRAN 90 running on an HP 9000 workstation to conduct the search for the optimal refueling patterns for a CANDU reactor at the equilibrium state. The results showed that it was possible to apply genetic algorithms to automatically search for the refueling channels of the CANDU reactor. The optimal refueling patterns were compared with the solutions obtained from the AUTOREFUEL program and the results were consistent with each other. (authors)« less

Discontinuous Mode Power Supply

NASA Technical Reports Server (NTRS)

Lagadinos, John; Poulos, Ethel

2012-01-01

A document discusses the changes made to a standard push-pull inverter circuit to avoid saturation effects in the main inverter power supply. Typically, in a standard push-pull arrangement, the unsymmetrical primary excitation causes variations in the volt second integral of each half of the excitation cycle that could lead to the establishment of DC flux density in the magnetic core, which could eventually cause saturation of the main inverter transformer. The relocation of the filter reactor normally placed across the output of the power supply solves this problem. The filter reactor was placed in series with the primary circuit of the main inverter transformer, and is presented as impedance against the sudden changes on the input current. The reactor averaged the input current in the primary circuit, avoiding saturation of the main inverter transformer. Since the implementation of the described change, the above problem has not reoccurred, and failures in the main power transistors have been avoided.

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

NASA Technical Reports Server (NTRS)

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

1985-01-01

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

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

ERIC Educational Resources Information Center

General Accounting Office, Washington, DC.

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

Extrapolation techniques applied to matrix methods in neutron diffusion problems

NASA Technical Reports Server (NTRS)

Mccready, Robert R

1956-01-01

A general matrix method is developed for the solution of characteristic-value problems of the type arising in many physical applications. The scheme employed is essentially that of Gauss and Seidel with appropriate modifications needed to make it applicable to characteristic-value problems. An iterative procedure produces a sequence of estimates to the answer; and extrapolation techniques, based upon previous behavior of iterants, are utilized in speeding convergence. Theoretically sound limits are placed on the magnitude of the extrapolation that may be tolerated. This matrix method is applied to the problem of finding criticality and neutron fluxes in a nuclear reactor with control rods. The two-dimensional finite-difference approximation to the two-group neutron fluxes in a nuclear reactor with control rods. The two-dimensional finite-difference approximation to the two-group neutron-diffusion equations is treated. Results for this example are indicated.

Off-design temperature effects on nuclear fuel pins for an advanced space-power-reactor concept

NASA Technical Reports Server (NTRS)

Bowles, K. J.

1974-01-01

An exploratory out-of-reactor investigation was made of the effects of short-time temperature excursions above the nominal operating temperature of 990 C on the compatibility of advanced nuclear space-power reactor fuel pin materials. This information is required for formulating a reliable reactor safety analysis and designing an emergency core cooling system. Simulated uranium mononitride (UN) fuel pins, clad with tungsten-lined T-111 (Ta-8W-2Hf) showed no compatibility problems after heating for 8 hours at 2400 C. At 2520 C and above, reactions occurred in 1 hour or less. Under these conditions free uranium formed, redistributed, and attacked the cladding.

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

PubMed

Ballinger, J R

2010-11-01

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

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

PubMed Central

Ballinger, J R

2010-01-01

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

Observation of nuclear reactors on satellites with a balloon-borne gamma-ray telescope

NASA Technical Reports Server (NTRS)

O'Neill, Terrence J.; Kerrick, Alan D.; Ait-Ouamer, Farid; Tumer, O. Tumay; Zych, Allen D.

1989-01-01

Four Soviet nuclear-powered satellites flying over a double Compton gamma-ray telescope resulted in the detection of gamma rays with 0.3-8.0 MeV energies on April 15, 1988, as the balloonborne telescope searched, from a 35-km altitude, for celestial gamma-ray sources. The satellites included Cosmos 1900 and 1932. The USSR is the only nation currently employing moderated nuclear reactors for satellite power; reactors in space may cause significant problems for gamma-ray astronomy by increasing backgrounds, especially in the case of gamma-ray bursts.

Chemical Engineers Go to the Movies (Stimulating Problems for the Contemporary Undergraduate Student)

ERIC Educational Resources Information Center

Smart, Jimmy L.

2007-01-01

In this article, the author presents five problems that are representative of some of the "movie problems" that he has used on tests in various courses, including reactor design, heat transfer, mass transfer, engineering economics, and fluid mechanics. These problems tend to be open-ended. They can be challenging and can often be worked a variety…

Reactor experiments to study luminescence of He-Ne and He-Kr gaseous mixtures, excited by the products of 6Li (n, α) 3H nuclear reaction

NASA Astrophysics Data System (ADS)

Batyrbekov, E. G.; Gordienko, Yu. N.; Barsukov, N. I.; Ponkratov, Yu. V.; Kulsartov, T. V.; Khassenov, M. U.; Zaurbekova, Zh. A.; Tulubayev, Ye. Y.; Samarkhanov, K. K.

2018-04-01

The spectral studies of optical radiation of gaseous mixtures are of interest for solving problems associated with finding gaseous media with high energy conversion efficiency of nuclear reactions into the energy of laser or spontaneous emission [1, 2]. Such media can be used to extract energy from nuclear and fusion reactors in the form of optical radiation, and also to control and adjust the nuclear reactors parameters. This paper presents the preliminary results of the reactor experiments to study the spectral-luminescent properties of gas mixtures (based on He, Ne and Kr noble gases) excited by the products of 6Li(n,α)3H nuclear reaction at different levels of the stationary power of the IVG.1M reactor.

VERA Core Simulator methodology for pressurized water reactor cycle depletion

DOE PAGES

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane; ...

2017-01-12

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

A brief history of design studies on innovative nuclear reactors

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2014-09-01

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

ISP33 standard problem on the PACTEL facility

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

Purhonen, H.; Kouhia, J.; Kalli, H.

ISP33 is the first OECD/NEA/CSNI standard problem related to VVER type of pressurized water reactors. The reference reactor of the PACTEL test facility, which was used to carry out the ISP33 experiment, is the VVER-440 reactor, two of which are located near the Finnish city of Loviisa. The objective of the ISP33 test was to study the natural circulation behaviour of VVER-440 reactors at different coolant inventories. Natural circulation was considered as a suitable phenomenon to focus on by the first VVER related ISP due to its importance in most accidents and transients. The behaviour of the natural circulation wasmore » expected to be different compared to Western type of PWRs as a result of the effect of horizontal steam generators and the hot leg loop seals. This ISP was conducted as a blind problem. The experiment was started at full coolant inventory. Single-phase natural circulation transported the energy from the core to the steam generators. The inventory was then reduced stepwise at about 900 s intervals draining 60 kg each time from the bottom of the downcomer. the core power was about 3.7% of the nominal value. The test was terminated after the cladding temperatures began to rise. ATHLET, CATHARE, RELAP5 (MODs 3, 2.5 and 2), RELAP4/MOD6, DINAMIKA and TECH-M4 codes were used in 21 pre- and 20 posttest calculations submitted for the ISP33.« less

Nuclear reactor power as applied to a space-based radar mission

NASA Technical Reports Server (NTRS)

Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Bloomfield, H.; Heller, J.

1988-01-01

A space-based radar mission and spacecraft are examined to determine system requirements for a 300 kWe space nuclear reactor power system. The spacecraft configuration and its orbit, launch vehicle, and propulsion are described. Mission profiles are addressed, and storage in assembly orbit is considered. Dynamics and attitude control and the problems of nuclear and thermal radiation are examined.

Disclosure of the oscillations in kinetics of the reactor pressure vessel steel damage at fast neutron intensity decreasing

NASA Astrophysics Data System (ADS)

Krasikov, E.; Nikolaenko, V.

2017-01-01

Fast neutron intensity influence on reactor materials radiation damage is a critically important question in the problem of the correct use of the accelerated irradiation tests data for substantiation of the materials workability in real irradiation conditions that is low neutron intensity. Investigations of the fast neutron intensity (flux) influence on radiation damage and experimental data scattering reveal the existence of non-monotonous sections in kinetics of the reactor pressure vessels (RPV) steel damage. Discovery of the oscillations as indicator of the self-organization processes presence give reasons for new ways searching on reactor pressure vessel (RPV) steel radiation stability increasing and attempt of the self-restoring metal elaboration. Revealing of the wavelike process in the form of non monotonous parts of the kinetics of radiation embrittlement testifies that periodic transformation of the structure take place. This fact actualizes the problem of more precise definition of the RPV materials radiation embrittlement mechanisms and gives reasons for search of the ways to manage the radiation stability (nanostructuring and so on to stimulate the radiation defects annihilation), development of the means for creating of more stableness self recovering smart materials.

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

Top shield temperatures, C and K Reactors

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

Agar, J.D.

1964-12-28

A modification program is now in progress at the C and K Reactors consisting of an extensive renovation of the graphite channels in the vertical safety rod ststems. The present VSR channels are being enlarged by a graphite coring operation and channel sleeves will be installed in the larger channels. One problem associated with the coring operation is the danger of damaging top thermal shield cooling tubes located close to the VSR channels to such an extent that these tubes will have to be removed from service. If such a condition should exist at one or a number of locationsmore » in the top shield of the reactors after reactor startup, the question remains -- what would the resulting temperatures be of the various components of the top shields? This study was initiated to determine temperature distributions in the top shield complex at the C and K Reactors for various top thermal shield coolant system conditions. Since the top thermal shield cooling system at C Reactor is different than those at the K Reactors, the study was conducted separately for the two different systems.« less

Pressurized-water reactor internals aging degradation study. Phase 1

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

Luk, K.H.

1993-09-01

This report documents the results of a Phase I study on the effects of aging degradations on pr internals. Primary stressers for internals an generated by the primary coolant flow in the they include unsteady hydrodynamic forces and pump-generated pressure pulsations. Other stressors are applied loads, manufacturing processes, impurities in the coolant and exposures to fast neutron fluxes. A survey of reported aging-related failure information indicates that fatigue, stress corrosion cracking (SCC) and mechanical wear are the three major aging-related degradation mechanisms for PWR internals. Significant reported failures include thermal shield flow-induced vibration problems, SCC in guide tube support pinsmore » and core support structure bolts, fatigue-induced core baffle water-jet impingement problems and excess wear in flux thimbles. Many of the reported problems have been resolved by accepted engineering practices. Uncertainties remain in the assessment of long-term neutron irradiation effects and environmental factors in high-cycle fatigue failures. Reactor internals are examined by visual inspections and the technique is access limited. Improved inspection methods, especially one with an early failure detection capability, can enhance the safety and efficiency of reactor operations.« less

The history and perspective of Romania-USA cooperation in the field of technologic transfer of TRIGA reactor concept

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

Ciocaanescu, M.; Ionescu, M.

1996-08-01

The cooperation between Romania and the USA in the field of technologic transfer of nuclear research reactor technology began with the steady state 14 MW{sub t} TRIGA reactor, installed at INR Pitesti, Romania. It is the first in the range of TRIGA reactors proposed as a materials testing reactor. The first criticality was reached in November 19, 1979 and first operation at 14 MW{sub t} level was in February 1980. The paper will present the short history of this cooperation and the perspective for a new cooperation for building a Nuclear Heating Plant using the TRIGA reactor concept for demonstrationmore » purpose. The energy crisis is a world-wide problem which affects each country in different ways because the resources and the consumption are unfairly distributed. World-wide research points out that the fossil fuel sources are not to be considered the main energy sources for the long term as they are limited.« less

Feasibility study Part I - Thermal hydraulic analysis of LEU target for {sup 99}Mo production in Tajoura reactor

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

Bsebsu, F.M.; Abotweirat, F.; Elwaer, S.

2008-07-15

The Renewable Energies and Water Desalination Research Center (REWDRC), Libya, will implement the technology for {sup 99}Mo isotope production using LEU foil target, to obtain new revenue streams for the Tajoura nuclear research reactor and desiring to serve the Libyan hospitals by providing the medical radioisotopes. Design information is presented for LEU target with irradiation device and irradiation Beryllium (Be) unit in the Tajoura reactor core. Calculated results for the reactor core with LEU target at different level of power are presented for steady state and several reactivity induced accident situations. This paper will present the steady state thermal hydraulicmore » design and transient analysis of Tajoura reactor was loaded with LEU foil target for {sup 99}Mo production. The results of these calculations show that the reactor with LEU target during the several cases of transient are in safe and no problems will occur. (author)« less

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

NASA Technical Reports Server (NTRS)

Moran, Robert P.

2013-01-01

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

Corrosion and Corrosion Control in Light Water Reactors

NASA Astrophysics Data System (ADS)

Gordon, Barry M.

2013-08-01

Serious corrosion problems have plagued the light water reactor (LWR) industry for decades. The complex corrosion mechanisms involved and the development of practical engineering solutions for their mitigation will be discussed in this article. After a brief overview of the basic designs of the boiling water reactor (BWR) and pressurized water reactor (PWR), emphasis will be placed on the general corrosion of LWR containments, flow-accelerated corrosion of carbon steel components, intergranular stress corrosion cracking (IGSCC) in BWRs, primary water stress corrosion cracking (PWSCC) in PWRs, and irradiation-assisted stress corrosion cracking (IASCC) in both systems. Finally, the corrosion future of both plants will be discussed as plants extend their period of operation for an additional 20 to 40 years.

Anaerobic digestion of olive mill wastewaters in biofilm reactors packed with granular activated carbon and "Manville" silica beads.

PubMed

Bertin, Lorenzo; Berselli, Sara; Fava, Fabio; Petrangeli-Papini, Marco; Marchetti, Leonardo

2004-01-01

Anaerobic digestion is one of the most promising technologies for disposing olive mill wastewaters (OMWs). The process is generally carried out in the conventional contact bioreactors, which however are often unable to efficiently remove OMW phenolic compounds, that therefore occur in the effluents. The possibility of mitigating this problem by employing an anaerobic OMW-digesting microbial consortium passively immobilized in column reactors packed with granular activated carbon (GAC) or "Manville" silica beads (SB) was here investigated. Under batch conditions, both GAC- and SB-packed-bed biofilm reactors exhibited OMW COD and phenolic compound removal efficiencies markedly higher (from 60% to 250%) than those attained in a parallel anaerobic dispersed growth reactor developed with the same inoculum; GAC-reactor exhibited COD and phenolic compound depletion yields higher by 62% and 78%, respectively, than those achieved with the identically configured SB-biofilm reactor. Both biofilm reactors also mediated an extensive OMW remediation under continuous conditions, where GAC-reactor was much more effective than the corresponding SB-one, and showed a tolerance to high and variable organic loads along with a volumetric productivity in terms of COD and phenolic compound removal significantly higher than those averagely displayed by most of the conventional and packed-bed laboratory-scale reactors previously proposed for the OMW digestion.

Laboratory Reactor for Processing Carbon-Containing Sludge

NASA Astrophysics Data System (ADS)

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

2016-10-01

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

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

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

None

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

Switching of High-Voltage Cable Lines with Shunt Reactors

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

Sheskin, E. B., E-mail: evgeniy.sheskin@gmail.com; Evdokunin, G. A.

2016-05-15

The problem of disconnecting high-voltage cable lines with shunt reactors by SF{sub 6} circuit breakers is discussed. In these schemes it is possible to have a significant aperiodic component of the circuit breaker current that can prevent opening of the breaker. The authors propose methods for application to cable transmission lines which they believe will be optimal for ensuring normal disconnects.

Study of the Dynamics of a Pressurized Reactor as Integral Unit for Naval Propulsion; STUDIO DELLA DINAMICA DI UN REATTORE PRESSURIZZATO AD UNITA INTEGRALE PER PROPULSIONE NAVALE

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

Gilardi, E.; Cimorelli, L.

1963-07-01

The dynamic behavior of an integrated, pressurizedwater reactor with natural circulation was investigated both by analog computer techniques and a simplified analytical approach. Hydraulic instabilities due to the core or riser were considered, as well as overall stability and problems arising from heavy sea conditions. (auth)

"Bonjour History": Materials and Ideas for Teaching Louisiana's Cajun History = "Bonjour l'Histoire": Quelques Idees, Deux ou Trois Activites, et Plusieurs Materiaux pour l'Enseignement de l'Histoire des Cadiens en Louisiane.

ERIC Educational Resources Information Center

Mounier, Brenda

The goals of this teacher's guidebook and videotape are designed to incorporate Acadian (Cajun) history into the 4th grade social studies curriculum and the 4th and 5th grade Louisiana 30-minute daily French programs and French immersion programs. Another goal is to create an awareness, appreciation, and understanding of Acadian history in…

Flight Control Design - Best Practices

DTIC Science & Technology

2000-12-01

n’était pas universellement disponible à l’époque. La première partie du rapport donne quelques exemples de problèmes de commandes de vol. Ils...pitch axis. We can infer a lesson learned in the form of design guidance for control allocation or priority. Rigorous analysis is required to define...flight excitation and data gathering manoeuvres are safe and are sufficient to produce the required information. BP9.5 Time must be allocated in the

A brief history of design studies on innovative nuclear reactors

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

Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com

2014-09-30

In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USAmore » and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.« less

Nuclear Reactor Physics

NASA Astrophysics Data System (ADS)

Stacey, Weston M.

2001-02-01

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

Growth and characterization of III-V epitaxial films

NASA Astrophysics Data System (ADS)

Tripathi, A.; Adamski, J.

1991-11-01

Investigations were conducted on the growth of epitaxial layers using an Organo Metallic Chemical Vapor Deposition technique of selected III-V materials which are potentially useful for photonics and microwave devices. RL/ERX's MOCVD machine was leak checked for safety. The whole gas handling plumbing system has been leak checked and the problems were reported to the manufacturer, CVD Equipment Corporation of Dear Park, NY. CVD Equipment Corporation is making an effort to correct these problems and also supply the part according to our redesign specifications. One of the main emphasis during this contract period was understanding the operating procedure and writing an operating manual for this MOCVD machine. To study the dynamic fluid flow in the vertical reactor of this MOCVD machine, an experimental apparatus was designed, tested, and put together. This study gave very important information on the turbulent gas flow patterns in this vertical reactor. The turbulent flow affects the epitaxial growth adversely. This study will also help in redesigning a vertical reactor so that the turbulent gas flow can be eliminated.

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

PubMed

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

2018-05-01

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

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

DOE PAGES

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

2016-10-01

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

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

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

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

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

Alloying of steel and graphite by hydrogen in nuclear reactor

NASA Astrophysics Data System (ADS)

Krasikov, E.

2017-02-01

In traditional power engineering hydrogen may be one of the first primary source of equipment damage. This problem has high actuality for both nuclear and thermonuclear power engineering. Study of radiation-hydrogen embrittlement of the steel raises the question concerning the unknown source of hydrogen in reactors. Later unexpectedly high hydrogen concentrations were detected in irradiated graphite. It is necessary to look for this source of hydrogen especially because hydrogen flakes were detected in reactor vessels of Belgian NPPs. As a possible initial hypothesis about the enigmatical source of hydrogen one can propose protons generation during beta-decay of free neutrons поскольку inasmuch as protons detected by researches at nuclear reactors as witness of beta-decay of free neutrons.

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

Design and fabrication of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolytic oil production in Bangladesh

NASA Astrophysics Data System (ADS)

Aziz, Mohammad Abdul; Al-khulaidi, Rami Ali; Rashid, MM; Islam, M. R.; Rashid, MAN

2017-03-01

In this research, a development and performance test of a fixed-bed batch type pyrolysis reactor for pilot scale pyrolysis oil production was successfully completed. The characteristics of the pyrolysis oil were compared to other experimental results. A solid horizontal condenser, a burner for furnace heating and a reactor shield were designed. Due to the pilot scale pyrolytic oil production encountered numerous problems during the plant’s operation. This fixed-bed batch type pyrolysis reactor method will demonstrate the energy saving concept of solid waste tire by creating energy stability. From this experiment, product yields (wt. %) for liquid or pyrolytic oil were 49%, char 38.3 % and pyrolytic gas 12.7% with an operation running time of 185 minutes.

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

Amitava Sarkar; James K. Neathery; Burtron H. Davis

A fundamental filtration study was started to investigate the separation of Fischer-Tropsch Synthesis (FTS) liquids from iron-based catalyst particles. Slurry-phase FTS in slurry bubble column reactor systems is the preferred mode of operation since the reaction is highly exothermic. Consequently, heavy wax products in one approach may be separated from catalyst particles before being removed from the reactor system. Achieving an efficient wax product separation from iron-based catalysts is one of the most challenging technical problems associated with slurry-phase iron-based FTS and is a key factor for optimizing operating costs. The separation problem is further compounded by attrition of ironmore » catalyst particles and the formation of ultra-fine particles.« less

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

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

Lee, Changho; Yang, Won Sik

This paper presents the methods and performance of the MC2 -3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2 -2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2 -3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400more » 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2 -3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.« less

Development of attrition resistant iron-based Fischer-Tropsch catalysts

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

NONE

2000-09-20

The Fischer-Tropsch (F-T) reaction provides a way of converting coal-derived synthesis gas (CO+H{sub 2}) to liquid fuels. Since the reaction is highly exothermic, one of the major problems in control of the reaction is heat removal. Recent work has shown that the use of slurry bubble column reactors (SBCRs) can largely solve this problem. The use of iron-based catalysts is attractive not only due to their low cost and ready availability, but also due to their high water-gas shift activity which makes it possible to use these catalysts with low H{sub 2}/CO ratios. However, a serious problem with use ofmore » Fe catalysts in a SBCR is their tendency to undergo attrition. This can cause fouling/plugging of downstream filters and equipment, makes the separation of catalyst from the oil/wax product very difficult if not impossible, and results a steady loss of catalyst from the reactor. The objective of this research is to develop robust iron-based Fischer-Tropsch catalysts that have suitable activity, selectivity and stability to be used in the slurry bubble column reactor. Specifically we aim to develop to: (1) improve the performance and preparation procedure of the high activity, high attrition resistant, high alpha iron-based catalysts synthesized at Hampton University (2) seek improvements in the catalyst performance through variations in process conditions, pretreatment procedures and/or modifications in catalyst preparation steps and (3) investigate the performance in a slurry reactor. The effort during the reporting period has been devoted to effects of pretreating procedures, using H{sub 2}, CO and syngas (H{sub 2}/CO = 0.67) as reductants, on the performance (activity, selectivity and stability with time) of a precipitated iron catalyst (100Fe/5Cu/4.2K/10SiO{sub 2} on a mass basis ) during F-T synthesis were studied in a fixed-bed reactor.« less

Preliminary study on aerobic granular biomass formation with aerobic continuous flow reactor

NASA Astrophysics Data System (ADS)

Yulianto, Andik; Soewondo, Prayatni; Handajani, Marissa; Ariesyady, Herto Dwi

2017-03-01

A paradigm shift in waste processing is done to obtain additional benefits from treated wastewater. By using the appropriate processing, wastewater can be turned into a resource. The use of aerobic granular biomass (AGB) can be used for such purposes, particularly for the processing of nutrients in wastewater. During this time, the use of AGB for processing nutrients more reactors based on a Sequencing Batch Reactor (SBR). Studies on the use of SBR Reactor for AGB demonstrate satisfactory performance in both formation and use. SBR reactor with AGB also has been applied on a full scale. However, the use use of SBR reactor still posses some problems, such as the need for additional buffer tank and the change of operation mode from conventional activated sludge to SBR. This gives room for further reactor research with the use of a different type, one of which is a continuous reactor. The purpose of this study is to compare AGB formation using continuous reactor and SBR with same operation parameter. Operation parameter are Organic Loading Rate (OLR) set to 2,5 Kg COD/m3.day with acetate as substrate, aeration rate 3 L/min, and microorganism from Hospital WWTP as microbial source. SBR use two column reactor with volumes 2 m3, and continuous reactor uses continuous airlift reactor, with two compartments and working volume of 5 L. Results from preliminary research shows that although the optimum results are not yet obtained, AGB can be formed on the continuous reactor. When compared with AGB generated by SBR, then the characteristics of granular diameter showed similarities, while the sedimentation rate and Sludge Volume Index (SVI) characteristics showed lower yields.

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

Kochunas, Brendan; Collins, Benjamin; Stimpson, Shane

This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclidemore » transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. Lastly, these results provide confidence in VERA-CS’s capability to perform high-fidelity calculations for practical PWR reactor problems.« less

Neutronic Reactor III

NASA Astrophysics Data System (ADS)

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

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

ANALYTICAL CHEMISTRY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING DECEMBER 31, 1961

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

None

1962-02-01

Research and development progress is reported on analytlcal instrumentation, dlssolver-solution analyses, special research problems, reactor projects analyses, x-ray and spectrochemical analyses, mass spectrometry, optical and electron microscopy, radiochemical analyses, nuclear analyses, inorganic preparations, organic preparations, ionic analyses, infrared spectral studies, anodization of sector coils for the Analog II Cyclotron, quality control, process analyses, and the Thermal Breeder Reactor Projects Analytical Chemistry Laboratory. (M.C.G.)

Steam generator for liquid metal fast breeder reactor

DOEpatents

Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

1985-01-01

Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

Updated Chemical Kinetics and Sensitivity Analysis Code

NASA Technical Reports Server (NTRS)

Radhakrishnan, Krishnan

2005-01-01

An updated version of the General Chemical Kinetics and Sensitivity Analysis (LSENS) computer code has become available. A prior version of LSENS was described in "Program Helps to Determine Chemical-Reaction Mechanisms" (LEW-15758), NASA Tech Briefs, Vol. 19, No. 5 (May 1995), page 66. To recapitulate: LSENS solves complex, homogeneous, gas-phase, chemical-kinetics problems (e.g., combustion of fuels) that are represented by sets of many coupled, nonlinear, first-order ordinary differential equations. LSENS has been designed for flexibility, convenience, and computational efficiency. The present version of LSENS incorporates mathematical models for (1) a static system; (2) steady, one-dimensional inviscid flow; (3) reaction behind an incident shock wave, including boundary layer correction; (4) a perfectly stirred reactor; and (5) a perfectly stirred reactor followed by a plug-flow reactor. In addition, LSENS can compute equilibrium properties for the following assigned states: enthalpy and pressure, temperature and pressure, internal energy and volume, and temperature and volume. For static and one-dimensional-flow problems, including those behind an incident shock wave and following a perfectly stirred reactor calculation, LSENS can compute sensitivity coefficients of dependent variables and their derivatives, with respect to the initial values of dependent variables and/or the rate-coefficient parameters of the chemical reactions.

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

NONE

During this time period, at WVU, the authors have obtained models for the kinetics of the HAS (higher alcohol synthesis) reaction over the Co-K-MoS{sub 2}/C catalyst. The Rotoberty reactor was then replaced in the reactor system by a plug-flow tubular reactor. Accordingly, the authors re-started the investigations on sulfide catalysts. The authors encountered and solved the leak problem from the sampling valve for the non-sulfided reactor system. They also modified the system to eliminate the condensation problem. Accordingly, they are continuing their kinetic studies on the reduced Mo-Ni-K/C catalysts. They have set up an apparatus for temperature-programmed reduction (TPR) studies,more » and have obtained some interesting results on TPR characterizations. At UCC, the complete characterization of selected catalysts has been started. The authors sent nine selected types of ZnO, Zn/CrO and Zn/Cr/MnO catalysts and supports for BET surface area, SEM, XRD and ICP. They also sent fresh and spent samples of the Engelhard Zn/CrO catalyst impregnated with 3 wt% potassium for ISS and XPS testing. In Task 2, work on the design and optimization portion of this task, as well as on the fuel testing, is completed. All funds have been expended and there are no personnel working on this project.« less

Extension of the Bgl Broad Group Cross Section Library

NASA Astrophysics Data System (ADS)

Kirilova, Desislava; Belousov, Sergey; Ilieva, Krassimira

2009-08-01

The broad group cross-section libraries BUGLE and BGL are applied for reactor shielding calculation using the DOORS package based on discrete ordinates method and multigroup approximation of the neutron cross-sections. BUGLE and BGL libraries are problem oriented for PWR or VVER type of reactors respectively. They had been generated by collapsing the problem independent fine group library VITAMIN-B6 applying PWR and VVER one-dimensional radial model of the reactor middle plane using the SCALE software package. The surveillance assemblies (SA) of VVER-1000/320 are located on the baffle above the reactor core upper edge in a region where geometry and materials differ from those of the middle plane and the neutron field gradient is very high which would result in a different neutron spectrum. That is why the application of the fore-mentioned libraries for the neutron fluence calculation in the region of SA could lead to an additional inaccuracy. This was the main reason to study the necessity for an extension of the BGL library with cross-sections appropriate for the SA region. Comparative analysis of the neutron spectra of the SA region calculated by the VITAMIN-B6 and BGL libraries using the two-dimensional code DORT have been done with purpose to evaluate the BGL applicability for SA calculation.

Hybrid parallel code acceleration methods in full-core reactor physics calculations

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

Courau, T.; Plagne, L.; Ponicot, A.

2012-07-01

When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadraturemore » required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)« less

Assessment and mitigation of power quality problems for PUSPATI TRIGA Reactor (RTP)

NASA Astrophysics Data System (ADS)

Zakaria, Mohd Fazli; Ramachandaramurthy, Vigna K.

2017-01-01

An electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality are necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. To reduce these disturbances, mitigation techniques have been proposed, such as to install passive harmonic filters to reduce harmonic distortion, dynamic voltage restorer (DVR) to reduce voltage disturbances and isolate all sensitive and critical loads.

Varied applications of a new maximum-likelihood code with complete covariance capability. [FERRET, for data adjustment

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

Schmittroth, F.

1978-01-01

Applications of a new data-adjustment code are given. The method is based on a maximum-likelihood extension of generalized least-squares methods that allow complete covariance descriptions for the input data and the final adjusted data evaluations. The maximum-likelihood approach is used with a generalized log-normal distribution that provides a way to treat problems with large uncertainties and that circumvents the problem of negative values that can occur for physically positive quantities. The computer code, FERRET, is written to enable the user to apply it to a large variety of problems by modifying only the input subroutine. The following applications are discussed:more » A 75-group a priori damage function is adjusted by as much as a factor of two by use of 14 integral measurements in different reactor spectra. Reactor spectra and dosimeter cross sections are simultaneously adjusted on the basis of both integral measurements and experimental proton-recoil spectra. The simultaneous use of measured reaction rates, measured worths, microscopic measurements, and theoretical models are used to evaluate dosimeter and fission-product cross sections. Applications in the data reduction of neutron cross section measurements and in the evaluation of reactor after-heat are also considered. 6 figures.« less

Gas-phase optical fiber photocatalytic reactors for indoor air application: a preliminary study on performance indicators

NASA Astrophysics Data System (ADS)

Palmiste, Ü.; Voll, H.

2017-10-01

The development of advanced air cleaning technologies aims to reduce building energy consumption by reduction of outdoor air flow rates while keeping the indoor air quality at an acceptable level by air cleaning. Photocatalytic oxidation is an emerging technology for gas-phase air cleaning that can be applied in a standalone unit or a subsystem of a building mechanical ventilation system. Quantitative information on photocatalytic reactor performance is required to evaluate the technical and economic viability of the advanced air cleaning by PCO technology as an energy conservation measure in a building air conditioning system. Photocatalytic reactors applying optical fibers as light guide or photocatalyst coating support have been reported as an approach to address the current light utilization problems and thus, improve the overall efficiency. The aim of the paper is to present a preliminary evaluation on continuous flow optical fiber photocatalytic reactors based on performance indicators commonly applied for air cleaners. Based on experimental data, monolith-type optical fiber reactor performance surpasses annular-type optical fiber reactors in single-pass removal efficiency, clean air delivery rate and operating cost efficiency.

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

Alternative nuclear technologies

NASA Astrophysics Data System (ADS)

Schubert, E.

1981-10-01

The lead times required to develop a select group of nuclear fission reactor types and fuel cycles to the point of readiness for full commercialization are compared. Along with lead times, fuel material requirements and comparative costs of producing electric power were estimated. A conservative approach and consistent criteria for all systems were used in estimates of the steps required and the times involved in developing each technology. The impact of the inevitable exhaustion of the low- or reasonable-cost uranium reserves in the United States on the desirability of completing the breeder reactor program, with its favorable long-term result on fission fuel supplies, is discussed. The long times projected to bring the most advanced alternative converter reactor technologies the heavy water reactor and the high-temperature gas-cooled reactor into commercial deployment when compared to the time projected to bring the breeder reactor into equivalent status suggest that the country's best choice is to develop the breeder. The perceived diversion-proliferation problems with the uranium plutonium fuel cycle have workable solutions that can be developed which will enable the use of those materials at substantially reduced levels of diversion risk.

Nuclear radiation problems, unmanned thermionic reactor ion propulsion spacecraft

NASA Technical Reports Server (NTRS)

Mondt, J. F.; Sawyer, C. D.; Nakashima, A.

1972-01-01

A nuclear thermionic reactor as the electric power source for an electric propulsion spacecraft introduces a nuclear radiation environment that affects the spacecraft configuration, the use and location of electrical insulators and the science experiments. The spacecraft is conceptually configured to minimize the nuclear shield weight by: (1) a large length to diameter spacecraft; (2) eliminating piping penetrations through the shield; and (3) using the mercury propellant as gamma shield. Since the alumina material is damaged by the high nuclear radiation environment in the reactor it is desirable to locate the alumina insulator outside the reflector or develop a more radiation resistant insulator.

Optimization of post-column reactor radius in capillary high performance liquid chromatography Effect of chromatographic column diameter and particle diameter

PubMed Central

Xu, Hongjuan; Weber, Stephen G.

2006-01-01

A post-column reactor consisting of a simple open tube (Capillary Taylor Reactor) affects the performance of a capillary LC in two ways: stealing pressure from the column and adding band spreading. The former is a problem for very small radius reactors, while the latter shows itself for large reactor diameters. We derived an equation that defines the observed number of theoretical plates (Nobs) taking into account the two effects stated above. Making some assumptions and asserting certain conditions led to a final equation with a limited number of variables, namely chromatographic column radius, reactor radius and chromatographic particle diameter. The assumptions and conditions are that the van Deemter equation applies, the mass transfer limitation is for intraparticle diffusion in spherical particles, the velocity is at the optimum, the analyte’s retention factor, k′, is zero, the post-column reactor is only long enough to allow complete mixing of reagents and analytes and the maximum operating pressure of the pumping system is used. Optimal ranges of the reactor radius (ar) are obtained by comparing the number of observed theoretical plates (and theoretical plates per time) with and without a reactor. Results show that the acceptable reactor radii depend on column diameter, particle diameter, and maximum available pressure. Optimal ranges of ar become narrower as column diameter increases, particle diameter decreases or the maximum pressure is decreased. When the available pressure is 4000 psi, a Capillary Taylor Reactor with 12 μm radius is suitable for all columns smaller than 150 μm (radius) packed with 2–5 μm particles. For 1 μm packing particles, only columns smaller than 42.5 μm (radius) can be used and the reactor radius needs to be 5 μm. PMID:16494886

A review of the Los Alamos effort in the development of nuclear rocket propulsion

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

Durham, F.P.; Kirk, W.L.; Bohl, R.J.

1991-01-01

This paper reviews the achievements of the Los Alamos nuclear rocket propulsion program and describes some specific reactor design and testing problems encountered during the development program along with the progress made in solving these problems. The relevance of these problems to a renewed nuclear thermal rocket development program for the Space Exploration Initiative (SEI) is discussed. 11 figs.

Flow rate analysis of wastewater inside reactor tanks on tofu wastewater treatment plant

NASA Astrophysics Data System (ADS)

Mamat; Sintawardani, N.; Astuti, J. T.; Nilawati, D.; Wulan, D. R.; Muchlis; Sriwuryandari, L.; Sembiring, T.; Jern, N. W.

2017-03-01

The research aimed to analyse the flow rate of the wastewater inside reactor tanks which were placed a number of bamboo cutting. The resistance of wastewater flow inside reactor tanks might not be occurred and produce biogas fuel optimally. Wastewater from eleven tofu factories was treated by multi-stages anaerobic process to reduce its organic pollutant and produce biogas. Biogas plant has six reactor tanks of which its capacity for waste water and gas dome was 18 m3 and 4.5 m3, respectively. Wastewater was pumped from collecting ponds to reactors by either serial or parallel way. Maximum pump capacity, head, and electrical motor power was 5m3/h, 50m, and 0.75HP, consecutively. Maximum pressure of biogas inside the reactor tanks was 55 mbar higher than atmosphere pressure. A number of 1,400 pieces of cutting bamboo at 50-60 mm diameter and 100 mm length were used as bacteria growth media inside each reactor tank, covering around 14,287 m2 bamboo area, and cross section area of inner reactor was 4,9 m2. In each reactor, a 6 inches PVC pipe was installed vertically as channel. When channels inside reactor were opened, flow rate of wastewater was 6x10-1 L.sec-1. Contrary, when channels were closed on the upper part, wastewater flow inside the first reactor affected and increased gas dome. Initially, wastewater flowed into each reactor by a gravity mode with head difference between the second and third reactor was 15x10-2m. However, head loss at the second reactor was equal to the third reactor by 8,422 x 10-4m. As result, wastewater flow at the second and third reactors were stagnant. To overcome the problem pump in each reactor should be installed in serial mode. In order to reach the output from the first reactor and the others would be equal, and biogas space was not filled by wastewater, therefore biogas production will be optimum.

French Regulatory practice and experience feedback on steam generator tube integrity

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

Sandon, G.

1997-02-01

This paper summarizes the way the French Safety Authority applies regulatory rules and practices to the problem of steam generator tube cracking in French PWR reactors. There are 54 reactors providing 80% of French electrical consumption. The Safety Authority closely monitors the performance of tubes in steam generators, and requires application of a program which deals with problems prior to the actual development of leakage. The actual rules regarding such performance are flexible, responding to the overall performance of operating steam generators. In addition there is an inservice inspection service to examine tubes during shutdown, and to monitor steam generatorsmore » for leakage during operation, with guidelines for when generators must be pulled off line.« less

Activation product transport in fusion reactors. [RAPTOR

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

Klein, A.C.

1983-01-01

Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. Similar problems are experienced around fission reactor systems. The determination of the transport of radioactive corrosion and neutron sputtering products through the system is achieved using the computer code RAPTOR. This code calculates the mass transfer of a number of activation products based on the corrosion and sputtering rates through the system, the depositionmore » and release characteristics of various plant components, the neturon flux spectrum, as well as other plant parameters. RAPTOR assembles a system of first order linear differential equations into a matrix equation based upon the reactor system parameters. Included in the transfer matrix are the deposition and erosion coefficients, and the decay and activation data for the various plant nodes and radioactive isotopes. A source vector supplies the corrosion and neutron sputtering source rates. This matrix equation is then solved using a matrix operator technique to give the specific activity distribution of each radioactive species throughout the plant. Once the amount of mass transfer is determined, the photon transport due to the radioactive corrosion and sputtering product sources can be evaluated, and dose rates around the plant components of interest as a function of time can be determined. This method has been used to estimate the radiation hazards around a number of fusion reactor system designs.« less

Safety and core design of large liquid-metal cooled fast breeder reactors

NASA Astrophysics Data System (ADS)

Qvist, Staffan Alexander

In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

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.

Sidération myocardique au cours d'une intoxication au monoxyde de carbone (CO) chez une femme enceinte

PubMed Central

Coulibaly, Mahamadoun; Berdai, Mohamed Adnane; Labib, Smael; Harandou, Mustapha

2015-01-01

L'intoxication au monoxyde de carbone (CO) est la première cause de décès par intoxication en France. La littérature est ancienne et peu connue. Les signes les plus fréquents de l'intoxication sont la triade: Céphalées; asthénie, faiblesse musculaire surtout des membres inférieurs. Ses conséquences sont potentiellement graves pour le fœtus quand elle survient chez la femme enceinte, il est particulièrement exposé au risque d'hypoxie en raison de la forte affinité de son hémoglobine pour le CO qui traverse aisément le placenta. Les événements cardiovasculaires ne sont pas rares et peuvent être responsable d'une morbi-mortalité assez importante qui peuvent être d'apparition rapide ou secondaire mais régressent habituellement en quelques jours. Des SCA peuvent survenir lors d'une une intoxication au CO avec à l'extrême infarctus myocardique avec surélévation du segment ST. Il paraît légitime de proposer pour toutes les patientes: l’éloignement maternel de la source de CO; l'oxygénothérapie à 100% au masque facial par les services de secours et pendant le transfert; le traitement par oxygénothérapie hyperbare pour toutes les femmes enceintes, le plus rapidement possible et quelque soit l’âge gestationnel. PMID:26405502

ADVANCED DESIGNS OF MAGNETIC JACK-TYPE CONTROL ROD DRIVE

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

Young, J.N.

1959-11-01

The magnetic jack is a device for positioning the control rods In a nuclear reactor, especially in a reactor containing water under pressure. Magnetic actuation precludes the need for shaft seals and eliminates the problems associated with mechanisms operating in water. It consists of a pressure shell, four sets of external stationary magnet coils (hold, grip, lift, pull down), and one Internal moving part (ammature) that impants linear motion to a cluster of rods. (W.L.H.)

Laser Boron Fusion Reactor With Picosecond Petawatt Block Ignition

NASA Astrophysics Data System (ADS)

Hora, Heinrich; Eliezer, Shalom; Wang, Jiaxiang; Korn, Georg; Nissim, Noaz; Xu, Yan-Xia; Lalousis, Paraskevas; Kirchhoff, Gotz J.; Miley, George H.

2018-05-01

For developing a laser boron fusion reactor driven by picosecond laser pulses of more than 30 petawatts power, advances are reported about computations for the plasma block generation by the dielectric explosion of the interaction. Further results are about the direct drive ignition mechanism by a single laser pulse without the problems of spherical irradiation. For the sufficiently large stopping lengths of the generated alpha particles in the plasma results from other projects can be used.

Synthesis of Multi-Walled Carbon Nanotubes from Plastic Waste Using a Stainless-Steel CVD Reactor as Catalyst.

PubMed

Tripathi, Pranav K; Durbach, Shane; Coville, Neil J

2017-09-22

The disposal of non-biodegradable plastic waste without further upgrading/downgrading is not environmentally acceptable and many methods to overcome the problem have been proposed. Herein we indicate a simple method to make high-value nanomaterials from plastic waste as a partial solution to the environmental problem. Laboratory-based waste centrifuge tubes made of polypropylene were chosen as a carbon source to show the process principle. In the process, multi-walled carbon nanotubes (MWCNTs) were synthesized from plastic waste in a two-stage stainless steel 316 (SS 316) metal tube that acted as both reactor vessel and catalyst. The steel reactor contains Fe (and Ni, and various alloys), which act as the catalyst for the carbon conversion process. The reaction and products were studied using electron probe microanalysis, thermogravimetric analysis, Raman spectroscopy and transmission electron microscopy and scanning electron microscopy. Optimization studies to determine the effect of different parameters on the process showed that the highest yield and most graphitized MWCNTs were formed at 900 °C under the reaction conditions used (yield 42%; Raman I D / I G ratio = 0.48). The high quality and high yield of the MWCNTs that were produced in a flow reactor from plastic waste using a two stage SS 316 chemical vapor deposition (CVD) furnace did not require the use of an added catalyst.

Synthesis of Multi-Walled Carbon Nanotubes from Plastic Waste Using a Stainless-Steel CVD Reactor as Catalyst

PubMed Central

Durbach, Shane

2017-01-01

The disposal of non-biodegradable plastic waste without further upgrading/downgrading is not environmentally acceptable and many methods to overcome the problem have been proposed. Herein we indicate a simple method to make high-value nanomaterials from plastic waste as a partial solution to the environmental problem. Laboratory-based waste centrifuge tubes made of polypropylene were chosen as a carbon source to show the process principle. In the process, multi-walled carbon nanotubes (MWCNTs) were synthesized from plastic waste in a two-stage stainless steel 316 (SS 316) metal tube that acted as both reactor vessel and catalyst. The steel reactor contains Fe (and Ni, and various alloys), which act as the catalyst for the carbon conversion process. The reaction and products were studied using electron probe microanalysis, thermogravimetric analysis, Raman spectroscopy and transmission electron microscopy and scanning electron microscopy. Optimization studies to determine the effect of different parameters on the process showed that the highest yield and most graphitized MWCNTs were formed at 900 °C under the reaction conditions used (yield 42%; Raman ID/IG ratio = 0.48). The high quality and high yield of the MWCNTs that were produced in a flow reactor from plastic waste using a two stage SS 316 chemical vapor deposition (CVD) furnace did not require the use of an added catalyst. PMID:28937596

Significance of breeding in fast nuclear reactors

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

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

1983-12-01

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

Three dimensional contact/impact methodology

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

Kulak, R.F.

1987-01-01

The simulation of three-dimensional interface mechanics between reactor components and structures during static contact or dynamic impact is necessary to realistically evaluate their structural integrity to off-normal loads. In our studies of postulated core energy release events, we have found that significant structure-structure interactions occur in some reactor vessel head closure designs and that fluid-structure interactions occur within the reactor vessel. Other examples in which three-dimensional interface mechanics play an important role are: (1) impact response of shipping casks containing spent fuel, (2) whipping pipe impact on reinforced concrete panels or pipe-to-pipe impact after a pipe break, (3) aircraft crashmore » on secondary containment structures, (4) missiles generated by turbine failures or tornados, and (5) drops of heavy components due to lifting accidents. The above is a partial list of reactor safety problems that require adequate treatment of interface mechanics and are discussed in this paper.« less

Yttrium and rare earth stabilized fast reactor metal fuel

DOEpatents

Guon, Jerold; Grantham, LeRoy F.; Specht, Eugene R.

1992-01-01

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Materials and fabrication technology of modules intended for irradiation tests of blanket tritium-breeding zones in Russian fusion reactor projects

NASA Astrophysics Data System (ADS)

Kapychev, V.; Davydov, D.; Gorokhov, V.; Ioltukhovskiy, A.; Kazennov, Yu; Tebus, V.; Frolov, V.; Shikov, A.; Shishkov, N.; Kovalenko, V.; Shishkin, N.; Strebkov, Yu

2000-12-01

This paper surveys the modules and materials of blanket tritium-breeding zones developed in the Russian Federation for fusion reactors. Synthesis of lithium orthosilicate, metasilicate and aluminate, fabrication of ceramic pellets and pebbles and experimental reactor units are described. Results of tritium extraction kinetics under irradiation in a water-graphite reactor at a thermal neutron flux of 5×10 13 neutron/(s cm2) are considered. At the present time, development and fabrication of lithium orthosilicate-beryllium modules of the tritium-breeding zone (TBZ), have been carried out within the framework of the ITER and DEMO projects. Two modules containing orthosilicate pellets, porous beryllium and beryllium pebbles are suggested for irradiation tests in the temperature range of 350-700°C. Technical problems associated with manufacturing of the modules are discussed.

Nuclear fuels - Present and future

NASA Astrophysics Data System (ADS)

Olander, D.

2009-06-01

The important developments in nuclear fuels and their problems are reviewed and compared with the status of present light-water reactor fuels. The limitations of LWR fuels are reviewed with respect to important recent concerns, namely provision of outlet coolant temperatures high enough for use in H 2 production, destruction of plutonium to eliminate proliferation concerns, and burning of the minor actinides to reduce the waste repository heat load and long-term radiation hazard. In addition to current oxide-based fuel rod designs, the hydride fuel with liquid-metal thermal bonding of the fuel-cladding gap is covered. Finally, two of the most promising Generation IV reactor concepts, the very high temperature reactor and the sodium fast reactor, and the accompanying reprocessing technologies, aqueous-based UREX+1a and pyrometallurgical, are summarized. In all of the topics covered, the thermodynamics involved in the fuel's behavior under irradiation and in the reprocessing schemes are emphasized.

Peaceful Uses of Fusion

DOE R&D Accomplishments Database

Teller, E.

1958-07-03

Applications of thermonuclear energy for peaceful and constructive purposes are surveyed. Developments and problems in the release and control of fusion energy are reviewed. It is pointed out that the future of thermonuclear power reactors will depend upon the construction of a machine that produces more electric energy than it consumes. The fuel for thermonuclear reactors is cheap and practically inexhaustible. Thermonuclear reactors produce less dangerous radioactive materials than fission reactors and, when once brought under control, are not as likely to be subject to dangerous excursions. The interaction of the hot plasma with magnetic fields opens the way for the direct production of electricity. It is possible that explosive fusion energy released underground may be harnessed for the production of electricity before the same feat is accomplished in controlled fusion processes. Applications of underground detonations of fission devices in mining and for the enhancement of oil flow in large low-specific-yield formations are also suggested.

A Simple Global View of Fuel Burnup

NASA Astrophysics Data System (ADS)

Sekimoto, Hiroshi

2017-01-01

Reactor physics and fuel burnup are discussed in order to obtain a simple global view of the effects of nuclear reactor characteristics to fuel cycle system performance. It may provide some idea of free thinking and overall vision, though it is still a small part of nuclear energy system. At the beginning of this lecture, governing equations for nuclear reactors are presented. Since the set of these equations is so big and complicated, it is simplified by imposing some extreme conditions and the nuclear equilibrium equation is derived. Some features of future nuclear equilibrium state are obtained by solving this equation. The contribution of a nucleus charged into reactor core to the system performance indexes such as criticality is worth for understanding the importance of each nuclide. It is called nuclide importance and can be evaluated by using the equations adjoint to the nuclear equilibrium equation. Examples of some importances and their application to criticalily search problem are presented.

Flow instability in particle-bed nuclear reactors

NASA Technical Reports Server (NTRS)

Kerrebrock, J. L.; Kalamas, J.

1993-01-01

A three-dimensional model of the stability of the particle-bed reactor is presented, in which the fluid has mobility in three dimensions. The model accurately represents the stability at low Re numbers as well as the effects of the cold and hot frits and of the heat conduction and radiation in the particle bed. The model can be easily extended to apply to the cylindrical geometry of particle-bed reactors. Exemplary calculations are carried out, showing that a particle bed without a cold frit would be subject to instability if operated at the high-temperature ratios used for nuclear rockets and at power densities below about 4 MW/l; since the desired power density for such a reactor is about 40 MW/l, the operation at design exit temperature but at reduced power could be hazardous. Calculations show however that it might be possible to remove the instability problem by appropriate combinations of cold and hot frits.

Evaluation of coupling approaches for thermomechanical simulations

DOE PAGES

Novascone, S. R.; Spencer, B. W.; Hales, J. D.; ...

2015-08-10

Many problems of interest, particularly in the nuclear engineering field, involve coupling between the thermal and mechanical response of an engineered system. The strength of the two-way feedback between the thermal and mechanical solution fields can vary significantly depending on the problem. Contact problems exhibit a particularly high degree of two-way feedback between those fields. This paper describes and demonstrates the application of a flexible simulation environment that permits the solution of coupled physics problems using either a tightly coupled approach or a loosely coupled approach. In the tight coupling approach, Newton iterations include the coupling effects between all physics,more » while in the loosely coupled approach, the individual physics models are solved independently, and fixed-point iterations are performed until the coupled system is converged. These approaches are applied to simple demonstration problems and to realistic nuclear engineering applications. The demonstration problems consist of single and multi-domain thermomechanics with and without thermal and mechanical contact. Simulations of a reactor pressure vessel under pressurized thermal shock conditions and a simulation of light water reactor fuel are also presented. Here, problems that include thermal and mechanical contact, such as the contact between the fuel and cladding in the fuel simulation, exhibit much stronger two-way feedback between the thermal and mechanical solutions, and as a result, are better solved using a tight coupling strategy.« less

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

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

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

Design of MSR primary circuit with minimum pressure losses

NASA Astrophysics Data System (ADS)

Noga, Tomáš; Žitek, Pavel; Valenta, Václav

This article describes a design of a MSR primary circuit with minimum pressure losses. It includes a brief description of this type of a reactor and its integral layout, properties, purpose, etc. The objective of this paper is to define problems of pressure losses calculation and to design a proper device for a primary circuit of MSR reactor, including its basic dimensions. Thanks to this, it can become an initial project for a construction of a real piece of work. This is the main contribution of the carried out study. Of course, this article is not a detailed solution, but it points out facts and problems, which future designers may have to face. The further step of our work will be a reconstruction of the current experiment for a two-stage flowing.

Cracked shaft detection on large vertical nuclear reactor coolant pump

NASA Technical Reports Server (NTRS)

Jenkins, L. S.

1985-01-01

Due to difficulty and radiation exposure associated with examination of the internals of large commercial nuclear reactor coolant pumps, it is necessary to be able to diagnose the cause of an excessive vibration problem quickly without resorting to extensive trial and error efforts. Consequently, it is necessary to make maximum use of all available data to develop a consistent theory which locates the problem area in the machine. This type of approach was taken at Three Mile Island, Unit #1, in February 1984 to identify and locate the cause of a continuously climbing vibration level of the pump shaft. The data gathered necessitated some in-depth knowledge of the pump internals to provide proper interpretation and avoid misleading conclusions. Therefore, the raw data included more than just the vibration characteristics. Pertinent details of the data gathered is shown and is necessary and sufficient to show that the cause of the observed vibration problem could logically only be a cracked pump shaft in the shaft overhang below the pump bearing.

Behavior of U 3Si 2 Fuel and FeCrAl Cladding under Normal Operating and Accident Reactor Conditions

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

Gamble, Kyle Allan Lawrence; Hales, Jason Dean; Barani, Tommaso

2016-09-01

As part of the Department of Energy's Nuclear Energy Advanced Modeling and Simulation program, an Accident Tolerant Fuel High Impact Problem was initiated at the beginning of fiscal year 2015 to investigate the behavior of \\usi~fuel and iron-chromium-aluminum (FeCrAl) claddings under normal operating and accident reactor conditions. The High Impact Problem was created in response to the United States Department of Energy's renewed interest in accident tolerant materials after the events that occurred at the Fukushima Daiichi Nuclear Power Plant in 2011. The High Impact Problem is a multinational laboratory and university collaborative research effort between Idaho National Laboratory, Losmore » Alamos National Laboratory, Argonne National Laboratory, and the University of Tennessee, Knoxville. This report primarily focuses on the engineering scale research in fiscal year 2016 with brief summaries of the lower length scale developments in the areas of density functional theory, cluster dynamics, rate theory, and phase field being presented.« less

Beta-spectrum shapes of forbidden β decays

NASA Astrophysics Data System (ADS)

Kostensalo, Joel; Suhonen, Jouni

2018-03-01

The neutrinoless ββ decay of atomic nuclei continues to attract fervent interest due to its potential to confirm the possible Majorana nature of the neutrino, and thus the nonconservation of the lepton number. At the same time, the direct dark matter experiments are looking for weakly interacting massive particles (WIMPs) through their scattering on nuclei. The neutrino-oscillation experiments on reactor antineutrinos base their analyses on speculations of β-spectrum shapes of nuclear decays, thus leading to the notorious “reactor antineutrino anomaly.” In all these experimental efforts, one encounters the problem of β-spectrum shapes of forbidden β decays, either as unwanted backgrounds or unknown components in the analyses of data. In this work, the problem of spectrum shapes is discussed and illustrated with a set of selected examples. The relation of the β-spectrum shapes to the problem of the effective value of the weak axial-vector coupling strength gA and the enhancement of the axial-charge matrix element is also pointed out.

Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

NASA Astrophysics Data System (ADS)

Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

2017-09-01

The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. 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 energy. 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. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

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

Adamov, E.O.; Lebedev, V.A.; Kuznetsov, Yu.N.

Zheleznogorsk is situated near the territorial center -- Krasnoyarsk on the Yenisei river. Mining and chemical complex is the main industrial enterprise of the town, which has been constructed for generation and used for isolation of weapons-grade plutonium. Heat supply to the chemical complex and town at the moment is largely provided by nuclear co-generation plant (NCGP) on the basis of the ADEh-2 dual-purpose reactor, generating 430 Gcal/h of heat and, partially, by coal backup peak-load boiler houses. NCGP also provides 73% of electric power consumed. In line with agreements between Russia and USA on strategic arms reduction and phasingmore » out of weapons-grade plutonium production, decommissioning of the ADEh-2 reactor by 2000 is planned. Thus, a problem arises relative to compensation for electric and thermal power generation for the needs of the town and industrial enterprises, which is now supplied by the reactor. A nuclear power plant constructed on the same site as a substituting power source should be considered as the most practical option. Basic requirements to the reactor of substituting nuclear power plant are as follows. It is to be a new generation reactor on the basis of verified technologies, having an operating prototype optimal for underground siting and permitting utmost utilization of the available mining workings and those being disengaged. NCGP with the reactor is to be constructed in the time period required and is to become competitive with other possible power sources. Analysis has shown that the VK-300 simplified vessel-type boiling reactor meets the requirements made in the maximum extent. Its design is based on the experience of the VK-50 reactor operation for a period of 30 years in Dimitrovgrad (Russia) and allows for experience in the development of the SBWR type reactors. The design of the reactor is discussed.« less

Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

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

Bernard, J.A.

1989-09-01

This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power onmore » spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.« less

Microchannel Reactors for ISRU Applications

NASA Astrophysics Data System (ADS)

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

2005-02-01

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

C3I for Crisis, Emergency and Consequence Management (C3I pour la gestion des crises, des urgences et de leurs consequences)

DTIC Science & Technology

2009-05-01

Quelque soit le contexte, l’aide à la décision passe par une analyse en profondeur de trois (3) aspects importants interdépendants, à savoir le...information, including suggestions for reducing this burden to Department of Defense , Washington Headquarters Services, Directorate for Information...type de menace, nécessite en effet d’adopter une approche collective de la sécurité étendue à une coopération avec de multiples organisations civiles

Instabilités et chaos dans les oscillateurs paramétriques optiques

NASA Astrophysics Data System (ADS)

Amon, A.; Suret, P.; Bielawski, S.; Derozier, D.; Zemmouri, J.; Lefranc, M.; Nizette, M.; Erneux, T.

2004-11-01

Nous discutons quelques mécanismes d'instabilité récemment observés dans un oscillateur paramétrique optique (OPO) : d'une part des instabilités opto-thermiques où le système oscille autour des courbes de résonance d'un ou plusieurs modes, d'autre part des oscillations rapides résultant de l'interaction de plusieurs modes transverses. La première observation expérimentale de chaos déterministe dans un OPO est également présentée.

Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

NASA Astrophysics Data System (ADS)

Pauzi, Anas Muhamad; Cioncolini, Andrea; Iacovides, Hector

2015-04-01

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

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

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

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

Institute of Electrical and Electronics Engineers, Nuclear Science Symposium, 18th, and Nuclear Power Systems Symposium, 3rd, San Francisco, Calif., November 3-5, 1971, Proceedings.

NASA Technical Reports Server (NTRS)

1972-01-01

Potential advantages of fusion power reactors are discussed together with the protection of the public from radioactivity produced in nuclear power reactors, and the significance of tritium releases to the environment. Other subjects considered are biomedical instrumentation, radiation damage problems, low level environmental radionuclide analysis systems, nuclear techniques in environmental research, nuclear instrumentation, and space and plasma instrumentation. Individual items are abstracted in this issue.

Investigation of Chirality Selection Mechanism of Single Walled Carbon Nanotube-3

DTIC Science & Technology

2017-12-14

however, several universal and intrinsic problems remain. First, since the dewetting of a thin catalyst film into particles upon heating is a... heated to 800 °C in 15 minutes under Ar atmosphere, maintained for various times, and cooled down to room temperature. - Annealing of Fe-implanted...located 12 cm downstream from the middle of the tube reactor. Then the reactor was heated to 820 °C over 15 min with flowing Ar gas. During the ramping

Hot zero power reactor calculations using the Insilico code

DOE PAGES

Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; ...

2016-03-18

In this paper we describe the reactor physics simulation capabilities of the insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that the insilico SP N solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various PWR problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

THE NUCLEAR RAMJET PROPULSION SYSTEM

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

Merkle, T.C.

1959-06-30

The most practical nuclear ramjet systems consist of a suituble inlet diffusor system followed by a singlepass, straight-through heat exchanger (reactor) which couples into a typical exhaust nozzle. Within this framework, possibilities ars governed by the aerodynamic requirements of flight, the nuclear requirements of the reactor, the chemical problems associated with breathing air, and the mechanical properties of materials at elevated temperatures. The major research and development areas which must be entered in the actual production of such an engine are discussed. (W.D.M.)

Development of the ageing management database of PUSPATI TRIGA reactor

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

Ramli, Nurhayati, E-mail: nurhayati@nm.gov.my; Tom, Phongsakorn Prak; Husain, Nurfazila

Since its first criticality in 1982, PUSPATI TRIGA Reactor (RTP) has been operated for more than 30 years. As RTP become older, ageing problems have been seen to be the prominent issues. In addressing the ageing issues, an Ageing Management (AgeM) database for managing related ageing matters was systematically developed. This paper presents the development of AgeM database taking into account all RTP major Systems, Structures and Components (SSCs) and ageing mechanism of these SSCs through the system surveillance program.

Anaerobic sequencing batch reactors for wastewater treatment: a developing technology.

PubMed

Zaiat, M; Rodrigues, J A; Ratusznei, S M; de Camargo, E F; Borzani, W

2001-01-01

This paper describes and discusses the main problems related to anaerobic batch and fed-batch processes for wastewater treatment. A critical analysis of the literature evaluated the industrial application viability and proposed alternatives to improve operation and control of this system. Two approaches were presented in order to make this anaerobic discontinuous process feasible for industrial application: (1) optimization of the operating procedures in reactors containing self-immobilized sludge as granules, and (2) design of bioreactors with inert support media for biomass immobilization.

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.

Solving phase appearance/disappearance two-phase flow problems with high resolution staggered grid and fully implicit schemes by the Jacobian-free Newton–Krylov Method

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

Zou, Ling; Zhao, Haihua; Zhang, Hongbin

2016-04-01

The phase appearance/disappearance issue presents serious numerical challenges in two-phase flow simulations. Many existing reactor safety analysis codes use different kinds of treatments for the phase appearance/disappearance problem. However, to our best knowledge, there are no fully satisfactory solutions. Additionally, the majority of the existing reactor system analysis codes were developed using low-order numerical schemes in both space and time. In many situations, it is desirable to use high-resolution spatial discretization and fully implicit time integration schemes to reduce numerical errors. In this work, we adapted a high-resolution spatial discretization scheme on staggered grid mesh and fully implicit time integrationmore » methods (such as BDF1 and BDF2) to solve the two-phase flow problems. The discretized nonlinear system was solved by the Jacobian-free Newton Krylov (JFNK) method, which does not require the derivation and implementation of analytical Jacobian matrix. These methods were tested with a few two-phase flow problems with phase appearance/disappearance phenomena considered, such as a linear advection problem, an oscillating manometer problem, and a sedimentation problem. The JFNK method demonstrated extremely robust and stable behaviors in solving the two-phase flow problems with phase appearance/disappearance. No special treatments such as water level tracking or void fraction limiting were used. High-resolution spatial discretization and second- order fully implicit method also demonstrated their capabilities in significantly reducing numerical errors.« less

Les cooperatives et l'electrification rurale du Quebec, 1945--1964

NASA Astrophysics Data System (ADS)

Dorion, Marie-Josee

Cette these est consacree a l'histoire de l'electrification rurale du Quebec, et, plus particulierement, a l'histoire des cooperatives d'electricite. Fondees par vagues successives a partir de 1945, les cooperatives rurales d'electricite ont ete actives dans plusieurs regions du Quebec et elles ont electrifie une partie significative des zones rurales. Afin de comprendre le contexte de la creation des cooperatives d'electricite, notre these debute (premiere partie) par une analyse du climat sociopolitique des annees precedant la naissance du systeme cooperatif d'electrification rurale. Nous y voyons de quelle facon l'electrification rurale devient progressivement, a partir de la fin des annees 1920, une question d'actualite a laquelle les divers gouvernements qui se succedent tentent de trouver une solution, sans engager---ou si peu---les fonds de l'Etat. En ce sens, la premiere etatisation et la mise sur pied d'Hydro-Quebec, en 1944, marquent une rupture quant au mode d'action privilegie jusque-la. La nouvelle societe d'Etat se voit cependant retirer son mandat d'electrifier le monde rural un an apres sa fondation, car le gouvernement Duplessis, de retour au pouvoir, prefere mettre en place son propre modele d'electrification rurale. Ce systeme repose sur des cooperatives d'electricite, soutenues par un organisme public, l'Office de l'electrification rurale (OER). L'OER suscite de grandes attentes de la part des ruraux et c'est par centaines qu'ils se manifestent. Cet engouement pour les cooperatives complique la tache de l'OER, qui doit superviser de nouvelles societes tout en assurant sa propre organisation. Malgre des hesitations et quelques delais introduits par un manque de connaissances techniques et de personnel qualifie, les commissaires de l'OER se revelent perspicaces et parviennent a mettre sur pied un systeme cooperatif d'electrification rurale qui produit des resultats rapides. Il leur faudra cependant compter sur l'aide des autres acteurs engages dans l'electrification, les organismes publics et les compagnies privees d'electricite. Cette periode de demarrage et d'organisation, traitee dans la deuxieme partie de la these, se termine en 1947-48, au moment ou l'OER et les cooperatives raffermissent leur maitrise du systeme cooperatif d'electrification rurale. Les annees 1948 a 1955 (troisieme partie de these) correspondent a une periode de croissance pour le mouvement cooperatif. Cette partie scrute ainsi le developpement des cooperatives, les vastes chantiers de construction et l'injection de millions de dollars dans l'electrification rurale. Cette troisieme partie prend egalement acte des premiers signes que quelque chose ne va pas si bien dans le monde cooperatif. Nous y verrons egalement les ruraux a l'oeuvre: comme membres, d'abord, mais aussi en tant que benevoles, puis a l'emploi des cooperatives. La quatrieme et derniere partie, les annees 1956 a 1964, aborde les changements majeurs qui ont cours dans l'univers cooperatif; il s'agit d'une ere nouvelle et difficile pour le mouvement cooperatif, dont les reseaux paraissent inadaptes aux changements de profil de la consommation d'electricite des usagers. L'OER sent alors le besoin de raffermir son controle des cooperatives, car il pressent les problemes et les defis auxquels elles auront a faire face. Notre etude se termine par l'acquisition des cooperatives par Hydro-Quebec, en 1963-64. Fondee sur des sources riches et variees, notre demarche propose un eclairage inedit sur une dimension importante de l'histoire de l'electricite au Quebec. Elle permet, ce faisant, de saisir les rouages et l'action de l'Etat sous un angle particulier, avant sa profonde transformation amorcee au cours des annees 1960. De meme, elle apporte quelques cles nouvelles pour une meilleure comprehension de la dynamique des milieux ruraux de cette periode.

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

Dinh, Nam; Athe, Paridhi; Jones, Christopher

The Virtual Environment for Reactor Applications (VERA) code suite is assessed in terms of capability and credibility against the Consortium for Advanced Simulation of Light Water Reactors (CASL) Verification and Validation Plan (presented herein) in the context of three selected challenge problems: CRUD-Induced Power Shift (CIPS), Departure from Nucleate Boiling (DNB), and Pellet-Clad Interaction (PCI). Capability refers to evidence of required functionality for capturing phenomena of interest while capability refers to the evidence that provides confidence in the calculated results. For this assessment, each challenge problem defines a set of phenomenological requirements against which the VERA software is assessed. Thismore » approach, in turn, enables the focused assessment of only those capabilities relevant to the challenge problem. The evaluation of VERA against the challenge problem requirements represents a capability assessment. The mechanism for assessment is the Sandia-developed Predictive Capability Maturity Model (PCMM) that, for this assessment, evaluates VERA on 8 major criteria: (1) Representation and Geometric Fidelity, (2) Physics and Material Model Fidelity, (3) Software Quality Assurance and Engineering, (4) Code Verification, (5) Solution Verification, (6) Separate Effects Model Validation, (7) Integral Effects Model Validation, and (8) Uncertainty Quantification. For each attribute, a maturity score from zero to three is assigned in the context of each challenge problem. The evaluation of these eight elements constitutes the credibility assessment for VERA.« less

Prudence and Technology

ERIC Educational Resources Information Center

Weinberg, Alvin M.

1971-01-01

Argues that perfected technology, not neo-Ludite response, is necessary for solution of world food and resource problems. Although energy supply will ultimately limit available food, reactors can supply sufficient power for 15 billion population. (AL)

Transport synthetic acceleration for long-characteristics assembly-level transport problems

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

Zika, M.R.; Adams, M.L.

2000-02-01

The authors apply the transport synthetic acceleration (TSA) scheme to the long-characteristics spatial discretization for the two-dimensional assembly-level transport problem. This synthetic method employs a simplified transport operator as its low-order approximation. Thus, in the acceleration step, the authors take advantage of features of the long-characteristics discretization that make it particularly well suited to assembly-level transport problems. The main contribution is to address difficulties unique to the long-characteristics discretization and produce a computationally efficient acceleration scheme. The combination of the long-characteristics discretization, opposing reflecting boundary conditions (which are present in assembly-level transport problems), and TSA presents several challenges. The authorsmore » devise methods for overcoming each of them in a computationally efficient way. Since the boundary angular data exist on different grids in the high- and low-order problems, they define restriction and prolongation operations specific to the method of long characteristics to map between the two grids. They implement the conjugate gradient (CG) method in the presence of opposing reflection boundary conditions to solve the TSA low-order equations. The CG iteration may be applied only to symmetric positive definite (SPD) matrices; they prove that the long-characteristics discretization yields an SPD matrix. They present results of the acceleration scheme on a simple test problem, a typical pressurized water reactor assembly, and a typical boiling water reactor assembly.« less

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233U fuel production: 1. Concept of a stationary Tokamak as a neutron source

NASA Astrophysics Data System (ADS)

Azizov, E. A.; Gladush, G. G.; Dokuka, V. N.; Khayrutdinov, R. R.

2015-12-01

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

Bio-charcoal production from municipal organic solid wastes

NASA Astrophysics Data System (ADS)

AlKhayat, Z. Q.

2017-08-01

The economic and environmental problems of handling the increasingly huge amounts of urban and/or suburban organic municipal solid wastes MSW, from collection to end disposal, in addition to the big fluctuations in power supply and other energy form costs for the various civilian needs, is studied for Baghdad city, the ancient and glamorous capital of Iraq, and a simple control device is suggested, built and tested by carbonizing these dried organic wastes in simple environment friendly bio-reactor in order to produce low pollution potential, economical and local charcoal capsules that might be useful for heating, cooking and other municipal uses. That is in addition to the solve of solid wastes management problem which involves huge human and financial resources and causes many lethal health and environmental problems. Leftovers of different social level residential campuses were collected, classified for organic materials then dried in order to be supplied into the bio-reactor, in which it is burnt and then mixed with small amounts of sugar sucrose that is extracted from Iraqi planted sugar cane, to produce well shaped charcoal capsules. The burning process is smoke free as the closed burner’s exhaust pipe is buried 1m underground hole, in order to use the subsurface soil as natural gas filter. This process has proved an excellent performance of handling about 120kg/day of classified MSW, producing about 80-100 kg of charcoal capsules, by the use of 200 l reactor volume.

Comparison of actinide production in traveling wave and pressurized water reactors

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

Osborne, A.G.; Smith, T.A.; Deinert, M.R.

The geopolitical problems associated with civilian nuclear energy production arise in part from the accumulation of transuranics in spent nuclear fuel. A traveling wave reactor is a type of breed-burn reactor that could, if feasible, reduce the overall production of transuranics. In one possible configuration, a cylinder of natural or depleted uranium would be subjected to a fast neutron flux at one end. The neutrons would transmute the uranium, producing plutonium and higher actinides. Under the right conditions, the reactor could become critical, at which point a self-stabilizing fission wave would form and propagate down the length of the reactormore » cylinder. The neutrons from the fission wave would burn the fissile nuclides and transmute uranium ahead of the wave to produce additional fuel. Fission waves in uranium are driven largely by the production and fission of {sup 239}Pu. Simulations have shown that the fuel burnup can reach values greater than 400 MWd/kgIHM, before fission products poison the reaction. In this work we compare the production of plutonium and minor actinides produced in a fission wave to that of a UOX fueled light water reactor, both on an energy normalized basis. The nuclide concentrations in the spent traveling wave reactor fuel are computed using a one-group diffusion model and are verified using Monte Carlo simulations. In the case of the pressurized water reactor, a multi-group collision probability model is used to generate the nuclide quantities. We find that the traveling wave reactor produces about 0.187 g/MWd/kgIHM of transuranics compared to 0.413 g/MWd/kgIHM for a pressurized water reactor running fuel enriched to 4.95 % and burned to 50 MWd/kgIHM. (authors)« less

A Theoretical Investigation of Oxidation Efficiency of a Volatile Removal Assembly Reactor Under Microgravity Conditions

NASA Technical Reports Server (NTRS)

Guo, Boyun

2005-01-01

Volatile Removal Assembly (VRA) is a subsystem of the Closed Environment Life Support System (CELSS) installed in the International Space Station. It is used for removing contaminants (volatile organics) in the wastewater produced by the space station crews. The major contaminants are formic acid, ethanol, and propylene glycol. The VRA contains a slim packbed reactor (3.5 cm diameter and four 28 cm long tubes in series) to perform catalyst oxidation of wastewater at elevated pressure and temperature under microgravity conditions. In the reactor, the contaminants are burned with oxygen gas (O2) to form water and carbon dioxide (CO2) that dissolves in the water stream. Optimal design of the reactor requires a thorough understanding about how the reactor performs under microgravity conditions. The objective of this study was to develop a mathematical model to interpret experimental data obtained from normal and microgravity conditions, and to predict the performance of VRA reactor under microgravity conditions. Catalyst oxidation kinetics and the total oxygen-water contact area control the efficiency of catalyst oxidation for mass transfer, which depends on oxygen gas holdup and distribution in the reactor. The process involves bubbly flow in porous media with chemical reactions in microgravity environment. This presents a unique problem in fluid dynamics that has not been studied. Guo et al. (2004) developed a mathematical model that predicts oxygen holdup in the VRA reactor. No mathematical model has been found in the literature that can be used to predict the efficiency of catalyst oxidation under microgravity conditions.

Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

PubMed

Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

2017-05-01

In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS 18 O/ 16 O monitoring for future method development is proposed.

Nuclear reactor spacer grid and ductless core component

DOEpatents

Christiansen, David W.; Karnesky, Richard A.

1989-01-01

The invention relates to a nuclear reactor spacer grid member for use in a liquid cooled nuclear reactor and to a ductless core component employing a plurality of these spacer grid members. The spacer grid member is of the egg-shell type and is constructed so that the walls of the cell members of the grid member are formed of a single thickness of metal to avoid tolerance problems. Within each cell member is a hydraulic spring which laterally constrains the nuclear material bearing rod which passes through each cell member against a hardstop in response to coolant flow through the cell member. This hydraulic spring is also suitable for use in a water cooled nuclear reactor. A core component constructed of, among other components, a plurality of these spacer grid members, avoids the use of a full length duct by providing spacer sleeves about the sodium tubes passing through the spacer grid members at locations between the grid members, thereby maintaining a predetermined space between adjacent grid members.

Materials for DEMO and reactor applications—boundary conditions and new concepts

NASA Astrophysics Data System (ADS)

Coenen, J. W.; Antusch, S.; Aumann, M.; Biel, W.; Du, J.; Engels, J.; Heuer, S.; Houben, A.; Hoeschen, T.; Jasper, B.; Koch, F.; Linke, J.; Litnovsky, A.; Mao, Y.; Neu, R.; Pintsuk, G.; Riesch, J.; Rasinski, M.; Reiser, J.; Rieth, M.; Terra, A.; Unterberg, B.; Weber, Th; Wegener, T.; You, J.-H.; Linsmeier, Ch

2016-02-01

DEMO is the name for the first stage prototype fusion reactor considered to be the next step after ITER towards realizing fusion. For the realization of fusion energy especially, materials questions pose a significant challenge already today. Heat, particle and neutron loads are a significant problem to material lifetime when extrapolating to DEMO. For many of the issues faced, advanced materials solutions are under discussion or already under development. In particular, components such as the first wall and the divertor of the reactor can benefit from introducing new approaches such as composites or new alloys into the discussion. Cracking, oxidation as well as fuel management are driving issues when deciding for new materials. Here {{{W}}}{{f}}/{{W}} composites as well as strengthened CuCrZr components together with oxidation resilient tungsten alloys allow the step towards a fusion reactor. In addition, neutron induced effects such as transmutation, embrittlement and after-heat and activation are essential. Therefore, when designing a component an approach taking into account all aspects is required.

Classification of Reactor Facility Operational State Using SPRT Methods with Radiation Sensor Networks

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

Ramirez Aviles, Camila A.; Rao, Nageswara S.

We consider the problem of inferring the operational state of a reactor facility by using measurements from a radiation sensor network, which is deployed around the facility’s ventilation stack. The radiation emissions from the stack decay with distance, and the corresponding measurements are inherently random with parameters determined by radiation intensity levels at the sensor locations. We fuse measurements from network sensors to estimate the intensity at the stack, and use this estimate in a one-sided Sequential Probability Ratio Test (SPRT) to infer the on/off state of the reactor facility. We demonstrate the superior performance of this method over conventionalmore » majority vote fusers and individual sensors using (i) test measurements from a network of NaI sensors, and (ii) emulated measurements using radioactive effluents collected at a reactor facility stack. We analytically quantify the performance improvements of individual sensors and their networks with adaptive thresholds over those with fixed ones, by using the packing number of the radiation intensity space.« less

Final report, PT IP-535-C: Test of smaller VSR`s in DR reactor

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

Vaughn, A.D.

1963-04-17

Because of rod-sticking problems at DR Reactor, a knuckle rod of B Reactor design was installed in vertical safety channel number 28. The substitute VSR, which has a smaller diameter than the original DR rod, has been tested for its operational feasibility including both drop time and reactivity effect. The reactivity effect of the rod was estimated by comparison of the reactivity transient caused by insertion of the specific B-type rod after scramming into the pile, with similar transients caused by normal vertical safety rod in an adjacent channel. This document lists the indicated relative control strength of the rodmore » as an empirical basis for future safety calculations. Results indicate that the B-type knuckel rod in DR reactor is about 80% as strong as a normal DR vertical safety rod if used in equivalent flux distribution and location; this makes it reasonable to assume that the local control strength of the B-type knuckel rod is 98 {mu}b.« less

Supercritical water oxidation - Microgravity solids separation

NASA Technical Reports Server (NTRS)

Killilea, William R.; Hong, Glenn T.; Swallow, Kathleen C.; Thomason, Terry B.

1988-01-01

This paper discusses the application of supercritical water oxidation (SCWO) waste treatment and water recycling technology to the problem of waste disposal in-long term manned space missions. As inorganic constituents present in the waste are not soluble in supercritical water, they must be removed from the organic-free supercritical fluid reactor effluent. Supercritical water reactor/solids separator designs capable of removing precipitated solids from the process' supercritical fluid in zero- and low- gravity environments are developed and evaluated. Preliminary experiments are then conducted to test the concepts. Feed materials for the experiments are urine, feces, and wipes with the addition of reverse osmosis brine, the rejected portion of processed hygiene water. The solid properties and their influence on the design of several oxidation-reactor/solids-separator configurations under study are presented.

Preliminary study of fusion reactor: Solution of Grad Shapranov equation

NASA Astrophysics Data System (ADS)

Setiawan, Y.; Fermi, N.; Su'ud, Z.

2012-06-01

Nuclear fussion is prospective energy sources for the future due to the abundance of the fuel and can be categorized and clean energy sources. The problem is how to contain very hot plasma of temperature few hundreed million degrees safety and reliably. Tokamax type fussion reactors is considered as the most prospective concept. To analyze the plasma confining process and its movement Grad-Shavranov equation must be solved. This paper discuss about solution of Grad-Shavranov equation using Whittaker function. The formulation is then applied to the ITER design and example.

Neutron physics with accelerators

NASA Astrophysics Data System (ADS)

Colonna, N.; Gunsing, F.; Käppeler, F.

2018-07-01

Neutron-induced nuclear reactions are of key importance for a variety of applications in basic and applied science. Apart from nuclear reactors, accelerator-based neutron sources play a major role in experimental studies, especially for the determination of reaction cross sections over a wide energy span from sub-thermal to GeV energies. After an overview of present and upcoming facilities, this article deals with state-of-the-art detectors and equipment, including the often difficult sample problem. These issues are illustrated at selected examples of measurements for nuclear astrophysics and reactor technology with emphasis on their intertwined relations.

No sleep in the deep for Russian subs

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

Handler, J.

1993-04-01

In the Russian Far East, dozens of nuclear-powered submarines, once a threat to Western navies, are now a threat to the environment. Between mid-1989 and 1993, over 80 Russian nuclear submarines were removed from service. Nearly 80 more will be retired by the year 2000. Most of these submarines contain two nuclear reactors. The many decommisioned submarines have overwhelmed the limited funds and processing capacity of the Russian Navy. Problems include removal of the fuel, scrapping of the submarines, and safe disposal of the radioactive reactor vessels.

Consolidated fuel reprocessing program

NASA Astrophysics Data System (ADS)

1985-04-01

A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

Chemistry experience in the primary heat transport circuits of Kraftwerk Union pressurized water reactors

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

Riess, R.

Chosen for this description of the selected Kraftwerk Union (KWU) pressurized water reactor units were Obrigheim (KWO, 345 MW(e)), Stade (KKS, 662 (MW(e)), Borselle (KCB, 477 MW(e)), and Biblis (KWB-A, 1204 MW(e)). The experience at these plants shows that with a special startup procedure and a proper chemical control of the primary heat transport system that influences general corrosion, selective types of corrosion, corrosion product activity transport and resulting contamination, and radiation-induced decomposition, KWU units have no basic problems.

Perspectives on multifield models

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

Banerjee, S.

1997-07-01

Multifield models for prediction of nuclear reactor thermalhydraulics are reviewed from the viewpoint of their structure and requirements for closure relationships. Their strengths and weaknesses are illustrated with examples, indicating that they are effective in predicting separated and distributed flow regimes, but have problems for flows with large oscillations. Needs for multifield models are also discussed in the context of reactor operations and accident simulations. The highest priorities for future developments appear to relate to closure relationships for three-dimensional multifield models with emphasis on those needed for calculations of phase separation and entrainment/de-entrainment in complex geometries.

Controlling Weapons-Grade Fissile Material

ERIC Educational Resources Information Center

Rotblat, J.

1977-01-01

Discusses the problems of controlling weapons-grade fissionable material. Projections of the growth of fission nuclear reactors indicates sufficient materials will be available to construct 300,000 atomic bombs each containing 10 kilograms of plutonium by 1990. (SL)

The diversity and unit of reactor noise theory

NASA Astrophysics Data System (ADS)

Kuang, Zhifeng

The study of reactor noise theory concerns questions about cause and effect relationships, and utilisation of random noise in nuclear reactor systems. The diversity of reactor noise theory arises from the variety of noise sources, the various mathematical treatments applied and various practical purposes. The neutron noise in zero- energy systems arises from the fluctuations in the number of neutrons per fission, the time between nuclear events, and the type of reactions. It can be used to evaluate system parameters. The mathematical treatment is based on the master equation of stochastic branching processes. The noise in power reactor systems is given rise by random processes of technological origin such as vibration of mechanical parts, boiling of the coolant, fluctuations of temperature and pressure. It can be used to monitor reactor behaviour with the possibility of detecting malfunctions at an early stage. The mathematical treatment is based on the Langevin equation. The unity of reactor noise theory arises from the fact that useful information from noise is embedded in the second moments of random variables, which lends the possibility of building up a unified mathematical description and analysis of the various reactor noise sources. Exploring such possibilities is the main subject among the three major topics reported in this thesis. The first subject is within the zero power noise in steady media, and we reported on the extension of the existing theory to more general cases. In Paper I, by use of the master equation approach, we have derived the most general Feynman- and Rossi-alpha formulae so far by taking the full joint statistics of the prompt and all the six groups of delayed neutron precursors, and a multiple emission source into account. The involved problems are solved with a combination of effective analytical techniques and symbolic algebra codes (Mathematica). Paper II gives a numerical evaluation of these formulae. An assessment of the contribution of the terms that are novel as compared to the traditional formulae has been made. The second subject treats a problem in power reactor noise with the Langevin formalism. With a very few exceptions, in all previous work the diffusion approximation was used. In order to extend the treatment to transport theory, in Paper III, we introduced a novel method, i.e. Padé approximation via Lanczos algorithm to calculate the transfer function of a finite slab reactor described by one-group transport equation. It was found that the local-global decomposition of the neutron noise, formerly only reproduced in at least 2- group theory, can be reconstructed. We have also showed the existence of a boundary layer of the neutron noise close to the boundary. Finally, we have explored the possibility of building up a unified theory to account for the coexistence of zero power and power reactor noise in a system. In Paper IV, a unified description of the neutron noise is given by the use of backward master equations in a model where the cross section fluctuations are given as a simple binary pseudorandom process. The general solution contains both the zero power and power reactor noise concurrently, and they can be extracted individually as limiting cases of the general solution. It justified the separate treatments of zero power and power reactor noise. The result was extended to the case including one group of delayed neutron precursors in Paper V.

Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

NASA Astrophysics Data System (ADS)

Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

2017-10-01

The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

Gas core reactors for actinide transmutation and breeder applications

NASA Technical Reports Server (NTRS)

Clement, J. D.; Rust, J. H.

1978-01-01

This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

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

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

Krieg, R.

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

Determine Operating Reactor to Use for the 2016 PCI Level 1 Milestone

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

Clarno, Kevin T.

2016-01-30

The Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) Level 1 milestone to “Assess the analysis capability for core-wide [pressurized water reactor] PWR Pellet- Clad Interaction (PCI) screening and demonstrate detailed 3-D analysis on selected sub-region” (L1:CASL.P13.03) requires a particular type of nuclear power plant for the assessment. This report documents the operating reactor and cycles chosen for this assessment in completion of the physics integration (PHI) milestone to “Determine Operating Reactor to use for PCI L1 Milestone” (L3:PHI.CMD.P12.02). Watts Bar Unit 1 experienced (at least) one fuel rod failure in each of cycles 6 and 7, andmore » at least one was deemed to be duty related rather than being primarily related to a manufacturing defect or grid effects. This brief report documents that the data required to model cycles 1–12 of Watts Bar Unit 1 using VERA-CS contains sufficient data to model the PHI portion of the PCI challenge problem. A list of additional data needs is also provided that will be important for verification and validation of the BISON results.« less

Development of Ultra-Fine Multigroup Cross Section Library of the AMPX/SCALE Code Packages

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

Jeon, Byoung Kyu; Sik Yang, Won; Kim, Kang Seog

The Consortium for Advanced Simulation of Light Water Reactors Virtual Environment for Reactor Applications (VERA) neutronic simulator MPACT is being developed by Oak Ridge National Laboratory and the University of Michigan for various reactor applications. The MPACT and simplified MPACT 51- and 252-group cross section libraries have been developed for the MPACT neutron transport calculations by using the AMPX and Standardized Computer Analyses for Licensing Evaluations (SCALE) code packages developed at Oak Ridge National Laboratory. It has been noted that the conventional AMPX/SCALE procedure has limited applications for fast-spectrum systems such as boiling water reactor (BWR) fuels with very highmore » void fractions and fast reactor fuels because of its poor accuracy in unresolved and fast energy regions. This lack of accuracy can introduce additional error sources to MPACT calculations, which is already limited by the Bondarenko approach for resolved resonance self-shielding calculation. To enhance the prediction accuracy of MPACT for fast-spectrum reactor analyses, the accuracy of the AMPX/SCALE code packages should be improved first. The purpose of this study is to identify the major problems of the AMPX/SCALE procedure in generating fast-spectrum cross sections and to devise ways to improve the accuracy. For this, various benchmark problems including a typical pressurized water reactor fuel, BWR fuels with various void fractions, and several fast reactor fuels were analyzed using the AMPX 252-group libraries. Isotopic reaction rates were determined by SCALE multigroup (MG) calculations and compared with continuous energy (CE) Monte Carlo calculation results. This reaction rate analysis revealed three main contributors to the observed differences in reactivity and reaction rates: (1) the limitation of the Bondarenko approach in coarse energy group structure, (2) the normalization issue of probability tables, and (3) neglect of the self-shielding effect of resonance-like cross sections at high energy range such as (n,p) cross section of Cl35. The first error source can be eliminated by an ultra-fine group (UFG) structure in which the broad scattering resonances of intermediate-weight nuclides can be represented accurately by a piecewise constant function. A UFG AMPX library was generated with modified probability tables and tested against various benchmark problems. The reactivity and reaction rates determined with the new UFG AMPX library agreed very well with respect to Monte Carlo Neutral Particle (MCNP) results. To enhance the lattice calculation accuracy without significantly increasing the computational time, performing the UFG lattice calculation in two steps was proposed. In the first step, a UFG slowing-down calculation is performed for the corresponding homogenized composition, and UFG cross sections are collapsed into an intermediate group structure. In the second step, the lattice calculation is performed for the intermediate group level using the condensed group cross sections. A preliminary test showed that the condensed library reproduces the results obtained with the UFG cross section library. This result suggests that the proposed two-step lattice calculation approach is a promising option to enhance the applicability of the AMPX/SCALE system to fast system analysis.« less

Tritium resources available for fusion reactors

NASA Astrophysics Data System (ADS)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future fusion reactors.

Decrire et enseigner une competence de communication: remarques sur quelques solutions de continuite. L'Enseignement de la competence de communication en langues secondes. (Describing and Teaching Communicative Competence: Some Remarks on Solutions of Continuity. The Teaching of Communicative Competence in Second Languages.) Acts of the Colloquium of the Swiss Interuniversity Commission for Applied Linguistics. CILA Bulletin.

ERIC Educational Resources Information Center

Coste, Daniel

Two projects of the Ecole Normale Superieure de Saint-Cloud (CREDIF) are described and critically analyzed in this paper: the definition of a threshold level, "Niveau-seuil," in French and a learning module, "Looking for Work," intended to teach necessary written French to migrant workers. The threshold level section is a…

Silent Ship Research Applications and Operation. Volume 2. Unclassified Papers. Proceedings of a Conference Held at SACLANTCEN on 2-4 October 1984

DTIC Science & Technology

1985-01-15

moyennes calculees sur 62 bateaux sont priesnteesdans le tableau suivznt aoy’?nno mowvuw. desma yonn des mayavw dom % aupentation genral: bate"u A bateaux 5...i coque en bois, acier ou polyester. Le decoupaqe des variables en classes pernet de bitir deux matrices *un - tableau disjonctif complet", *un...pr6sentents quelques expertises ac oustiques provenant de deux t~tudes lr~alis~es .par le G.E.R.B.A.M. .I *la premiý-re, sur 95 thoniers ligneurs

The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

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

Hollaway, W.R.

1991-08-01

If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issuemore » through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW{sub e} IFR capacity for every three MW{sub e} Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years).« less

Implications of Zircaloy creep and growth to light water reactor performance

NASA Astrophysics Data System (ADS)

Franklin, David G.; Adamson, Ronald B.

1988-10-01

Deformation of zirconium alloy components in nuclear reactors has been a concern since the decision of Admiral Rickover to use them in the US Navy submarine reactors. With the exception of the first few light water reactors (LWRs) most of the core structural materials have been fabricated from either Zircaloy-2 or Zircaloy-4. Performance of these alloys has been extremely good, even though the effects of irradiation on deformation magnitudes and mechanisms were not fully appreciated until extensive service and in-reactor tests were accomplished. Since the reactor components are designed to operate at stress levels well below yield for normal conditions, the only significant deformation is time dependent. Although creep was anticipated, the enhancement by neutron irradiation and the stress-free, nearly constant-volume shape change known as irradiation growth were not known prior to materials testing in reactors under controlled conditions. Both of these phenomena have significant impact on performance and must be accounted for properly in design. Although irradiation creep and growth have resulted in only one significant performance problem (creep collapse of fuel cladding, which has been eliminated), deformation magnitudes and, particularly, differentials in strain magnitudes, are a continuing source of interest. Factors that affect dimensional stability due to both creep and growth include temperature, fluence, residual stress, texture, and microstructure. The first two are reactor variables and the others are related to component fabrication history. This paper includes a review of the applications of Zircaloy creep and growth to LWR fuel designs, a review of the impact of in-reactor creep and growth on fuel rod and fuel assembly performance, and comments on potential improvements. Since the reactor design, fuel design and the core environment in BWRs and PWRs are quite different, appropriate separation of the application of effects are made; of course, the basic phenomena are the same in both systems.

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

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

Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.

2015-12-15

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket basedmore » on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.« less

Optimality of affine control system of several species in competition on a sequential batch reactor

NASA Astrophysics Data System (ADS)

Rodríguez, J. C.; Ramírez, H.; Gajardo, P.; Rapaport, A.

2014-09-01

In this paper, we analyse the optimality of affine control system of several species in competition for a single substrate on a sequential batch reactor, with the objective being to reach a given (low) level of the substrate. We allow controls to be bounded measurable functions of time plus possible impulses. A suitable modification of the dynamics leads to a slightly different optimal control problem, without impulsive controls, for which we apply different optimality conditions derived from Pontryagin principle and the Hamilton-Jacobi-Bellman equation. We thus characterise the singular trajectories of our problem as the extremal trajectories keeping the substrate at a constant level. We also establish conditions for which an immediate one impulse (IOI) strategy is optimal. Some numerical experiences are then included in order to illustrate our study and show that those conditions are also necessary to ensure the optimality of the IOI strategy.

FY16 Status Report on NEAMS Neutronics Activities

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

Lee, C. H.; Shemon, E. R.; Smith, M. A.

2016-09-30

The goal of the NEAMS neutronics effort is to develop a neutronics toolkit for use on sodium-cooled fast reactors (SFRs) which can be extended to other reactor types. The neutronics toolkit includes the high-fidelity deterministic neutron transport code PROTEUS and many supporting tools such as a cross section generation code MC 2-3, a cross section library generation code, alternative cross section generation tools, mesh generation and conversion utilities, and an automated regression test tool. The FY16 effort for NEAMS neutronics focused on supporting the release of the SHARP toolkit and existing and new users, continuing to develop PROTEUS functions necessarymore » for performance improvement as well as the SHARP release, verifying PROTEUS against available existing benchmark problems, and developing new benchmark problems as needed. The FY16 research effort was focused on further updates of PROTEUS-SN and PROTEUS-MOCEX and cross section generation capabilities as needed.« less

Jet pump-drive system for heat removal

NASA Technical Reports Server (NTRS)

French, James R. (Inventor)

1987-01-01

The invention does away with the necessity of moving parts such as a check valve in a nuclear reactor cooling system. Instead, a jet pump, in combination with a TEMP, is employed to assure safe cooling of a nuclear reactor after shutdown. A main flow exists for a reactor coolant. A point of withdrawal is provided for a secondary flow. A TEMP, responsive to the heat from said coolant in the secondary flow path, automatically pumps said withdrawn coolant to a higher pressure and thus higher velocity compared to the main flow. The high velocity coolant is applied as a driver flow for the jet pump which has a main flow chamber located in the main flow circulation pump. Upon nuclear shutdown and loss of power for the main reactor pumping system, the TEMP/jet pump combination continues to boost the coolant flow in the direction it is already circulating. During the decay time for the nuclear reactor, the jet pump keeps running until the coolant temperature drops to a lower and safe temperature where the heat is no longer a problem. At this lower temperature, the TEMP/jet pump combination ceases its circulation boosting operation. When the nuclear reactor is restarted and the coolant again exceeds the lower temperature setting, the TEMP/jet pump automatically resumes operation. The TEMP/jet pump combination is thus automatic, self-regulating and provides an emergency pumping system free of moving parts.

Inverse modeling approach for evaluation of kinetic parameters of a biofilm reactor using tabu search.

PubMed

Kumar, B Shiva; Venkateswarlu, Ch

2014-08-01

The complex nature of biological reactions in biofilm reactors often poses difficulties in analyzing such reactors experimentally. Mathematical models could be very useful for their design and analysis. However, application of biofilm reactor models to practical problems proves somewhat ineffective due to the lack of knowledge of accurate kinetic models and uncertainty in model parameters. In this work, we propose an inverse modeling approach based on tabu search (TS) to estimate the parameters of kinetic and film thickness models. TS is used to estimate these parameters as a consequence of the validation of the mathematical models of the process with the aid of measured data obtained from an experimental fixed-bed anaerobic biofilm reactor involving the treatment of pharmaceutical industry wastewater. The results evaluated for different modeling configurations of varying degrees of complexity illustrate the effectiveness of TS for accurate estimation of kinetic and film thickness model parameters of the biofilm process. The results show that the two-dimensional mathematical model with Edward kinetics (with its optimum parameters as mu(max)rho(s)/Y = 24.57, Ks = 1.352 and Ki = 102.36) and three-parameter film thickness expression (with its estimated parameters as a = 0.289 x 10(-5), b = 1.55 x 10(-4) and c = 15.2 x 10(-6)) better describes the biofilm reactor treating the industry wastewater.

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.

PT-IP-759, channel caulking tests: C Reactor

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

Cooke, J.P.; Russell, A.

1965-03-19

The graphite movement which has occurred at the various reactors has been characterized by two problems: (1) Crooked channels and (2) cracks and miscellaneous voids where pieces of blocks are missing. Of these problems, the cracks and voids have been the most serious in the case of ball drops. Alleviation of the crooked channels can sometimes be accomplished by graphite removal methods such as broaching, but unless some method is found to prevent the balls from entering cracks, the total effect of a ball drop would still be intolerable. Of the two methods of closing the cracks, a paste caulkingmore » procedure is anticipated to be less expensive than sleeving, both in terms of cost of the operation and the number of process tube channels which might be lost. If the VSR channel does not require drastic straightening or entry of large tooling, satisfactory caulking can be done without removal of the step plug. ``Poison`` chain may be considered as an alternative to caulking or sleeving for those outer VSR channels where the sole use of balls is for ``total control`` rather than ``speed of control.`` The objectives of this test are (1) to authorize the experimental crack filling of one or two of the VSR channels at C Reactor with a wet mixture of graphite and sugar, (2) to demonstrate the durability of this mixture in subsequent normal reactor operation, and (3) to demonstrate by testing (actual or simulated ball drops) and borescoping, that the channels are or are not again acceptable for use with the normal charge of balls.« less

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

Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my; Cioncolini, Andrea; Iacovides, Hector

The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software calledmore » FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.« less

From Confrontation to Cooperation: 8th International Seminar on Nuclear War

NASA Astrophysics Data System (ADS)

Zichichi, A.; Dardo, M.

1992-09-01

The Table of Contents for the full book PDF is as follows: * OPENING SESSION * A. Zichichi: Opening Statements * R. Nicolosi: Opening Statements * MESSAGES * CONTRIBUTIONS * "The Contribution of the Erice Seminars in East-West-North-South Scientific Relations" * 1. LASER TECHNOLOGY * "Progress in laser technology" * "Progress in laboratory high gain ICF: prospects for the future" * "Applications of laser in metallurgy" * "Laser tissue interactions in medicine and surgery" * "Laser fusion" * "Compact X-ray lasers in the laboratory" * "Alternative method for inertial confinement" * "Laser technology in China" * 2. NUCLEAR AND CHEMICAL SAFETY * "Reactor safety and reactor design" * "Thereotical analysis and numerical modelling of heat transfer and fuel migration in underlying soils and constructive elements of nuclear plants during an accident release from the core" * "How really to attain reactor safely" * "The problem of chemical weapons" * "Long terms genetic effects of nuclear and chemical accidents" * "Features of the brain which are of importance in understanding the mode of operation of toxic substances and of radiation" * "CO2 and ultra safe reactors" * 3. USE OF MISSILES * "How to convert INF technology for peaceful scientific purposes" * "Beating words into plowshares: a proposal for the peaceful uses of retired nuclear warheads" * "Some thoughts on the peaceful use of retired nuclear warheads" * "Status of the HEFEST project" * 4. OZONE * "Status of the ozone layer problem" * 5. CONVENTIONAL AND NUCLEAR FORCE RESTRUCTURING IN EUROPE * 6. CONFLICT AVOIDANCE MODEL * 7. GENERAL DISCUSSION OF THE WORLD LAB PROJECTS * "East-West-North-South Collaboration in Subnuclear Physics" * "Status of the World Lab in the USSR" * CLOSING SESSION

Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

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

Bernnat, W.; Buck, M.; Mattes, M.

The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code ormore » memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)« less

View of Pakistan Atomic Energy Commission towards SMPR's in the light of KANUPP performance

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

Huseini, S.D.

1985-01-01

The developing countries in general do not have grid capacities adequate enough to incorporate standard size, economic but rather large nuclear power plants for maximum advantage. Therefore, small and medium size reactors (SMPR) have been and still are, of particular interest to the developing countries in spite of certain known problems with these reactors. Pakistan Atomic Energy Commission (PAEC) has been operating a CANDU type of a small PHWR plant since 1971 when it was connected to the local Karachi grid. This paper describes PAEC's view in the light of KANUPP performance with respect to such factors associated with SMPR'smore » as selection of suitable reactor size and type, its operation in a grid of small capacity, flexibility of operation and its role as a reliable source of electrical power.« less

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.

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

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

Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly casesmore » are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.« less

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

Owens, J.J.; Nejedlik, J.F.; Vogt, J.W.

The SNAP II system consists of a reactor heat source, a mercury Rankine engine, and an alternator. The problems involved in selecting materials for the SNAP II mercury system were studied. A discussion is given of the corrosion mechanisms involved in a system in which mercury is the working fluid. The problem resolves itself into selecting materials with the best combination of engineering properties for the application and highest resistance to mercury corrosion at the anticipated temperature. (auth)

Neutrino masses, neutrino oscillations, and cosmological implications

NASA Technical Reports Server (NTRS)

Stecker, F. W.

1982-01-01

Theoretical concepts and motivations for considering neutrinos having finite masses are discussed and the experimental situation on searches for neutrino masses and oscillations is summarized. The solar neutrino problem, reactor, deep mine and accelerator data, tri decay experiments and double beta-decay data are considered and cosmological implications and astrophysical data relating to neutrino masses are reviewed. The neutrino oscillation solution to the solar neutrino problem, the missing mass problem in galaxy halos and galaxy cluster galaxy formation and clustering, and radiative neutrino decay and the cosmic ultraviolet background radiation are examined.

Sensor placement in nuclear reactors based on the generalized empirical interpolation method

NASA Astrophysics Data System (ADS)

Argaud, J.-P.; Bouriquet, B.; de Caso, F.; Gong, H.; Maday, Y.; Mula, O.

2018-06-01

In this paper, we apply the so-called generalized empirical interpolation method (GEIM) to address the problem of sensor placement in nuclear reactors. This task is challenging due to the accumulation of a number of difficulties like the complexity of the underlying physics and the constraints in the admissible sensor locations and their number. As a result, the placement, still today, strongly relies on the know-how and experience of engineers from different areas of expertise. The present methodology contributes to making this process become more systematic and, in turn, simplify and accelerate the procedure.

Monte Carlo PDF method for turbulent reacting flow in a jet-stirred reactor

NASA Astrophysics Data System (ADS)

Roekaerts, D.

1992-01-01

A stochastic algorithm for the solution of the modeled scalar probability density function (PDF) transport equation for single-phase turbulent reacting flow is described. Cylindrical symmetry is assumed. The PDF is represented by ensembles of N representative values of the thermochemical variables in each cell of a nonuniform finite-difference grid and operations on these elements representing convection, diffusion, mixing and reaction are derived. A simplified model and solution algorithm which neglects the influence of turbulent fluctuations on mean reaction rates is also described. Both algorithms are applied to a selectivity problem in a real reactor.

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.

The optimization of nuclear power plants operation modes in emergency situations

NASA Astrophysics Data System (ADS)

Zagrebayev, A. M.; Trifonenkov, A. V.; Ramazanov, R. N.

2018-01-01

An emergency situations resulting in the necessity for temporary reactor trip may occur at the nuclear power plant while normal operating mode. The paper deals with some of the operation c aspects of nuclear power plant operation in emergency situations and during threatened period. The xenon poisoning causes limitations on the variety of statements of the problem of calculating characteristics of a set of optimal reactor power off controls. The article show a possibility and feasibility of new sets of optimization tasks for the operation of nuclear power plants under conditions of xenon poisoning in emergency circumstances.

Two Stage Anaerobic Reactor Design and Treatment To Produce Biogas From Mixed Liquor of Vegetable Waste

NASA Astrophysics Data System (ADS)

Budiastuti, H.; Ghozali, M.; Wicaksono, H. K.; Hadiansyah, R.

2018-01-01

Municipal solid waste has become a common challenged problem to be solved for developing countries including Indonesia. Municipal solid waste generating is always bigger than its treatment to reduce affect of environmental pollution. This research tries to contribute to provide an alternative solution to treat municipal solid waste to produce biogas. Vegetable waste was obtained from Gedebage Market, Bandung and starter as a source of anaerobic microorganisms was cow dung obtained from a cow farm in Lembang. A two stage anaerobic reactor was designed and built to treat the vegetable waste in a batch run. The capacity of each reactor is 20 liters but its active volume in each reactor is 15 liters. Reactor 1 (R1) was fed up with mixture of filtered blended vegetable waste and water at ratio of 1:1 whereas Reactor 2 (R2) was filled with filtered mixed liquor of cow dung and water at ratio of 1:1. Both mixtures were left overnight before use. Into R1 it was added EM-4 at concentration of 10%. pH in R1 was maintained at 5 - 6.5 whereas pH in R1 was maintained at 6.5 - 7.5. Temperature of reactors was not maintained to imitate the real environmental temperature. Parameters taken during experiment were pH, temperature, COD, MLVSS, and composition of biogas. The performance of reactor built was shown from COD efficiencies reduction obtained of about 60% both in R1 and R2, pH average in R1 of 4.5 ± 1 and R2 of 7 ± 0.6, average temperature in both reactors of 25 ± 2°C. About 1L gas produced was obtained during the last 6 days of experiment in which CH4 obtained was 8.951 ppm and CO2 of 1.087 ppm. The maximum increase of MLVSS in R1 reached 156% and R2 reached 89%.

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

NASA Astrophysics Data System (ADS)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

CFD Study of Full-Scale Aerobic Bioreactors: Evaluation of Dynamic O2 Distribution, Gas-Liquid Mass Transfer and Reaction

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

Humbird, David; Sitaraman, Hariswaran; Stickel, Jonathan

If advanced biofuels are to measurably displace fossil fuels in the near term, they will have to operate at levels of scale, efficiency, and margin unprecedented in the current biotech industry. For aerobically-grown products in particular, scale-up is complex and the practical size, cost, and operability of extremely large reactors is not well understood. Put simply, the problem of how to attain fuel-class production scales comes down to cost-effective delivery of oxygen at high mass transfer rates and low capital and operating costs. To that end, very large reactor vessels (>500 m3) are proposed in order to achieve favorable economiesmore » of scale. Additionally, techno-economic evaluation indicates that bubble-column reactors are more cost-effective than stirred-tank reactors in many low-viscosity cultures. In order to advance the design of extremely large aerobic bioreactors, we have performed computational fluid dynamics (CFD) simulations of bubble-column reactors. A multiphase Euler-Euler model is used to explicitly account for the spatial distribution of air (i.e., gas bubbles) in the reactor. Expanding on the existing bioreactor CFD literature (typically focused on the hydrodynamics of bubbly flows), our simulations include interphase mass transfer of oxygen and a simple phenomenological reaction representing the uptake and consumption of dissolved oxygen by submerged cells. The simulations reproduce the expected flow profiles, with net upward flow in the center of column and downward flow near the wall. At high simulated oxygen uptake rates (OUR), oxygen-depleted regions can be observed in the reactor. By increasing the gas flow to enhance mixing and eliminate depleted areas, a maximum oxygen transfer (OTR) rate is obtained as a function of superficial velocity. These insights regarding minimum superficial velocity and maximum reactor size are incorporated into NREL's larger techno-economic models to supplement standard reactor design equations.« less

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

Approaches to Deal with Irradiated Graphite in Russia - Proposal for New IAEA CRP on Graphite Waste Management - 12364

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

Kascheev, Vladimir; Poluektov, Pavel; Ustinov, Oleg

The problems of spent reactor graphite are being shown, the options of its disposal is considered. Burning method is selected as the most efficient and waste-free. It is made a comparison of amounts of {sup 14}C that entering the environment in a natural way during the operation of nuclear power plants (NPPs) and as a result of the proposed burning of spent reactor graphite. It is shown the possibility of burning graphite with the arrival of {sup 14}C into the atmosphere within the maximum allowable emissions. This paper analyzes the different ways of spent reactor graphite treatment. It is shownmore » the possibility of its reprocessing by burning method in the air flow. It is estimated the effect of this technology to the overall radiation environment and compared its contribution to the general background radiation due to cosmic radiation and NPPs emission. It is estimated the maximum permissible speeds of burning reactor graphite (for example, RBMK graphite) for areas with different conditions of agricultural activities. (authors)« 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

Numerical simulations of epitaxial growth process in MOVPE reactor as a tool for design of modern semiconductors for high power electronics

NASA Astrophysics Data System (ADS)

Skibinski, Jakub; Caban, Piotr; Wejrzanowski, Tomasz; Kurzydlowski, Krzysztof J.

2014-10-01

In the present study numerical simulations of epitaxial growth of gallium nitride in Metal Organic Vapor Phase Epitaxy reactor AIX-200/4RF-S is addressed. Epitaxial growth means crystal growth that progresses while inheriting the laminar structure and the orientation of substrate crystals. One of the technological problems is to obtain homogeneous growth rate over the main deposit area. Since there are many agents influencing reaction on crystal area such as temperature, pressure, gas flow or reactor geometry, it is difficult to design optimal process. According to the fact that it's impossible to determine experimentally the exact distribution of heat and mass transfer inside the reactor during crystal growth, modeling is the only solution to understand the process precisely. Numerical simulations allow to understand the epitaxial process by calculation of heat and mass transfer distribution during growth of gallium nitride. Including chemical reactions in numerical model allows to calculate the growth rate of the substrate and estimate the optimal process conditions for obtaining the most homogeneous product.

A bioreactor system for the nitrogen loop in a Controlled Ecological Life Support System

NASA Technical Reports Server (NTRS)

Saulmon, M. M.; Reardon, K. F.; Sadeh, W. Z.

1996-01-01

As space missions become longer in duration, the need to recycle waste into useful compounds rises dramatically. This problem can be addressed by the development of Controlled Ecological Life Support Systems (CELSS) (i.e., Engineered Closed/Controlled Eco-Systems (ECCES)), consisting of human and plant modules. One of the waste streams leaving the human module is urine. In addition to the reclamation of water from urine, recovery of the nitrogen is important because it is an essential nutrient for the plant module. A 3-step biological process for the recycling of nitrogenous waste (urea) is proposed. A packed-bed bioreactor system for this purpose was modeled, and the issues of reaction step segregation, reactor type and volume, support particle size, and pressure drop were addressed. Based on minimization of volume, a bioreactor system consisting of a plug flow immobilized urease reactor, a completely mixed flow immobilized cell reactor to convert ammonia to nitrite, and a plug flow immobilized cell reactor to produce nitrate from nitrite is recommended. It is apparent that this 3-step bioprocess meets the requirements for space applications.

Research on Liquid Management Technology in Water Tank and Reactor for Propulsion System with Hydrogen Production System Utilizing Aluminum and Water Reaction

NASA Astrophysics Data System (ADS)

Imai, Ryoji; Imamura, Takuya; Sugioka, Masatoshi; Higashino, Kazuyuki

2017-12-01

High pressure hydrogen produced by aluminum and water reaction is considered to be applied to space propulsion system. Water tank and hydrogen production reactor in this propulsion system require gas and liquid separation function under microgravity condition. We consider to install vane type liquid acquisition device (LAD) utilizing surface tension in the water tank, and install gas-liquid separation mechanism by centrifugal force which swirling flow creates in the hydrogen reactor. In water tank, hydrophilic coating was covered on both tank wall and vane surface to improve wettability. Function of LAD in water tank and gas-liquid separation in reaction vessel were evaluated by short duration microgravity experiments using drop tower facility. In the water tank, it was confirmed that liquid was driven and acquired on the outlet due to capillary force created by vanes. In addition of this, it was found that gas-liquid separation worked well by swirling flow in hydrogen production reactor. However, collection of hydrogen gas bubble was sometimes suppressed by aluminum alloy particles, which is open problem to be solved.

Reactor-released radionuclides in Susquehanna River sediments

USGS Publications Warehouse

Olsen, C.R.; Larsen, I.L.; Cutshall, N.H.; Donoghue, J.F.; Bricker, O.P.; Simpson, H.J.

1981-01-01

Three Mile Island (TMI) and Peach Bottom (PB) reactors have introduced 137Cs, 134Cs, 60Co, 58Co and several other anthropogenic radionuclides into the lower Susquehanna River. Here we present the release history for these nuclides (Table 1) and radionuclide concentration data (Table 2) for sediment samples collected in the river and upper portions of the Chesapeake Bay (Fig. 1) within a few months after the 28 March 1979 loss-of-coolant-water problem at TMI. Although we found no evidence for nuclides characteristic of a ruptured fuel element, we did find nuclides characteristic of routine operations. Despite the TMI incident, more than 95% of the total 134Cs input to the Susquehanna has been a result of controlled low-level releases from the PB site. 134Cs activity released into the river is effectively trapped by sediments with the major zones of reactor-nuclide accumulation behind Conowingo Dam and in the upper portions of Chesapeake Bay. The reported distributions document the fate of reactor-released radionuclides and their extent of environmental contamination in the Susquehanna-Upper Chesapeake Bay System. ?? 1981 Nature Publishing Group.

Technetium-99m production issues in the United Kingdom.

PubMed

Green, Christopher H

2012-04-01

Nuclear Medicine developed when it was realised that a radioisotopic substitution of Iodine-131 for the stable Iodine-127 would follow the same metabolic pathway in the body enabling the thyroid to be imaged and the thyroid uptake measured. The Iodine could be complexed with pharmaceutical substrates to enable other organs to be imaged, but its use was limited and high gamma energy and beta emission restricted the activity of each radiopharmaceutical used, leading to long acquisition times and degraded images. As a pure gamma emitter of 140 keV and with a 6-h half-life, Technetium-99m is a better radionuclide and images a wider range of bodily organs. However, its short half-life also requires it to be eluted from its mother radionuclide, Mo-99, in a generator, delivered weekly from radiopharmaceutical companies who obtain the Mo-99 in liquid form from high-flux research reactors. All went well till around 2007, when the NRU Reactor in Canada was closed and all other reactors went down for various periods for unrelated problems, leading to widespread Mo-99 shortages. Although the reactors have since recovered, they are 48 to 57 years old, and it seems that few governments have made any future provision such as building replacement reactors.

Hydrodynamics of high solids anaerobic reactor: Characterization of solid segregation and liquid mixing pattern in a pilot plant VALORGA facility under different reactor geometry.

PubMed

Álvarez, C; Colón, J; Lópes, A C; Fernández-Polanco, M; Benbelkacem, H; Buffière, P

2018-06-01

One of the main problems of dry anaerobic digestion plants treating urban solid waste is the loss of useful volume by the sedimentation of solids (inerts) into the bottom of the digester, or by accumulation of floating materials in its upper part. This entails a periodic cost of emptying and cleaning the digesters, a decrease in biogas production and complications in maintenance. Usually the sedimentation is a consequence of the heterogeneity of waste that, in addition to organic matter, drags particles of high density that end up obstructing the digesters. To reduce this bottleneck, URBASER has designed a new configuration of VALORGA reactor. That is, the VALORGA central wall has been removed and an inclined bottom has been added. To test the sedimentability and the overall performance of both configurations (current and new design), hydrodynamic tests have been carried out in a pilot digester (digester of 95 m 3 capacity). To simulate the liquid phase and the solid phase of the reactor, lithium tracers and tags of different densities with RFID (radio frequency identification reader) have been used respectively. The results of the study showed an improvement in the performance of the new reactor design at pilot level. Copyright © 2018 Elsevier Ltd. All rights reserved.

On heat loading, novel divertors, and fusion reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, M.; Valanju, P. M.; Mahajan, S. M.; Wiley, J. C.

2007-07-01

The limited thermal power handling capacity of the standard divertors (used in current as well as projected tokamaks) is likely to force extremely high (˜90%) radiation fractions frad in tokamak fusion reactors that have heating powers considerably larger than ITER [D. J. Campbell, Phys. Plasmas 8, 2041 (2001)]. Such enormous values of necessary frad could have serious and debilitating consequences on the core confinement, stability, and dependability for a fusion power reactor, especially in reactors with Internal Transport Barriers. A new class of divertors, called X-divertors (XD), which considerably enhance the divertor thermal capacity through a flaring of the field lines only near the divertor plates, may be necessary and sufficient to overcome these problems and lead to a dependable fusion power reactor with acceptable economics. X-divertors will lower the bar on the necessary confinement to bring it in the range of the present experimental results. Its ability to reduce the radiative burden imparts the X-divertor with a key advantage. Lower radiation demands allow sharply peaked density profiles that enhance the bootstrap fraction creating the possibility for a highly increased beta for the same beta normal discharges. The X-divertor emerges as a beta-enhancer capable of raising it by up to roughly a factor of 2.

The assembly and use of continuous flow systems for chemical synthesis.

PubMed

Britton, Joshua; Jamison, Timothy F

2017-11-01

The adoption of and opportunities in continuous flow synthesis ('flow chemistry') have increased significantly over the past several years. Continuous flow systems provide improved reaction safety and accelerated reaction kinetics, and have synthesised several active pharmaceutical ingredients in automated reconfigurable systems. Although continuous flow platforms are commercially available, systems constructed 'in-lab' provide researchers with a flexible, versatile, and cost-effective alternative. Herein, we describe the assembly and use of a modular continuous flow apparatus from readily available and affordable parts in as little as 30 min. Once assembled, the synthesis of a sulfonamide by reacting 4-chlorobenzenesulfonyl chloride with dibenzylamine in a single reactor coil with an in-line quench is presented. This example reaction offers the opportunity to learn several important skills including reactor construction, charging of a back-pressure regulator, assembly of stainless-steel syringes, assembly of a continuous flow system with multiple junctions, and yield determination. From our extensive experience of single-step and multistep continuous flow synthesis, we also describe solutions to commonly encountered technical problems such as precipitation of solids ('clogging') and reactor failure. Following this protocol, a nonspecialist can assemble a continuous flow system from reactor coils, syringes, pumps, in-line liquid-liquid separators, drying columns, back-pressure regulators, static mixers, and packed-bed reactors.

Wastewater treatment with submerged fixed bed biofilm reactor systems--design rules, operating experiences and ongoing developments.

PubMed

Schlegel, S; Koeser, H

2007-01-01

Wastewater treatment systems using bio-films that grow attached to a support media are an alternative to the widely used suspended growth activated sludge process. Different fixed growth biofilm reactors are commercially used for the treatment of municipal as well as industrial wastewater. In this paper a fairly new fixed growth biofilm system, the submerged fixed bed biofilm reactor (SFBBR), is discussed. SFBBRs are based on aerated submerged fixed open structured plastic media for the support of the biofilm. They are generally operated without sludge recirculation in order to avoid clogging of the support media and problems with the control of the biofilm. Reactor and process design considerations for these reactors are reviewed. Measures to ensure the development and maintenance of an active biofilm are examined. SFBBRs have been applied successfully to small wastewater treatment plants where complete nitrification but no high degree of denitrification is necessary. For the pre-treatment of industrial wastewater the use of SFBBRs is advantageous, especially in cases of wastewater with high organic loading or high content of compounds with low biodegradability. Performance data from exemplary commercial plants are given. Ongoing research and development efforts aim at achieving a high simultaneous total nitrogen (TN) removal of aerated SFBBRs and at improving the efficiency of TN removal in anoxic SFBBRs.

Technetium-99m production issues in the United Kingdom

PubMed Central

Green, Christopher H.

2012-01-01

Nuclear Medicine developed when it was realised that a radioisotopic substitution of Iodine-131 for the stable Iodine-127 would follow the same metabolic pathway in the body enabling the thyroid to be imaged and the thyroid uptake measured. The Iodine could be complexed with pharmaceutical substrates to enable other organs to be imaged, but its use was limited and high gamma energy and beta emission restricted the activity of each radiopharmaceutical used, leading to long acquisition times and degraded images. As a pure gamma emitter of 140 keV and with a 6-h half-life, Technetium-99m is a better radionuclide and images a wider range of bodily organs. However, its short half-life also requires it to be eluted from its mother radionuclide, Mo-99, in a generator, delivered weekly from radiopharmaceutical companies who obtain the Mo-99 in liquid form from high-flux research reactors. All went well till around 2007, when the NRU Reactor in Canada was closed and all other reactors went down for various periods for unrelated problems, leading to widespread Mo-99 shortages. Although the reactors have since recovered, they are 48 to 57 years old, and it seems that few governments have made any future provision such as building replacement reactors. PMID:22557795

Treatment of sanitary landfill leachates in a lab-scale gradual concentric chamber (GCC) reactor.

PubMed

Mendoza, Lourdes; Verstraete, Willy; Carballa, Marta

2010-03-01

Sanitary landfill leachates are a major environmental problem in South American countries where sanitary landfills are still constructed and appropriate designs for the treatment of these leachates remain problematic. The performance of a lab-scale Gradual Concentric Chamber (GCC) reactor for leachates treatment is presented in this study. Two types of sanitary landfill residuals were evaluated, one directly collected from the garbage trucks (JGL), with high organic strength (84 g COD/l) and the second one, a 6-month-generated leachate (YL) collected from the lagoon of the sanitary landfill in Quito, Ecuador, with an organic strength of 66 g COD/l. Different operational parameters, such as organic loading rate (OLR), temperature, recycling and aeration, were tested. The GCC reactor was found to be a robust technology to treat these high-strength streams with organic matter removal efficiencies higher than 65%. The best performance of the reactors (COD removal efficiencies of 75-80%) was obtained at a Hydraulic Retention Time (HRT) of around 20 h and at 35 degrees C, with an applied OLR up to 70 and 100 g COD/l per day. Overall, the GCC reactor concept appears worth to be further developed for the treatment of leachates in low-income countries.

The current state, main problems and directions in improving water chemistry at NPSs

NASA Astrophysics Data System (ADS)

Tyapkov, V. F.; Sharafutdinov, R. B.

2007-05-01

An analysis of the current state of managing water-chemistry (WC) at Russian nuclear power plants with type-VVER and-RBMK reactors presently in operation is presented. The main directions for improvement of WC are shown.

Diverse knowledges and competing interests: an essay on socio-technical problem-solving.

PubMed

di Norcia, Vincent

2002-01-01

Solving complex socio-technical problems, this paper claims, involves diverse knowledges (cognitive diversity), competing interests (social diversity), and pragmatism. To explain this view, this paper first explores two different cases: Canadian pulp and paper mill pollution and siting nuclear reactors in systematically sensitive areas of California. Solving such socio-technically complex problems involves cognitive diversity as well as social diversity and pragmatism. Cognitive diversity requires one to not only recognize relevant knowledges but also to assess their validity. Finally, it is suggested, integrating the resultant set of diverse relevant and valid knowledges determines the parameters of the solution space for the problem.

Membrane technology as a promising alternative in biodiesel production: a review.

PubMed

Shuit, Siew Hoong; Ong, Yit Thai; Lee, Keat Teong; Subhash, Bhatia; Tan, Soon Huat

2012-01-01

In recent years, environmental problems caused by the use of fossil fuels and the depletion of petroleum reserves have driven the world to adopt biodiesel as an alternative energy source to replace conventional petroleum-derived fuels because of biodiesel's clean and renewable nature. Biodiesel is conventionally produced in homogeneous, heterogeneous, and enzymatic catalysed processes, as well as by supercritical technology. All of these processes have their own limitations, such as wastewater generation and high energy consumption. In this context, the membrane reactor appears to be the perfect candidate to produce biodiesel because of its ability to overcome the limitations encountered by conventional production methods. Thus, the aim of this paper is to review the production of biodiesel with a membrane reactor by examining the fundamental concepts of the membrane reactor, its operating principles and the combination of membrane and catalyst in the catalytic membrane. In addition, the potential of functionalised carbon nanotubes to serve as catalysts while being incorporated into the membrane for transesterification is discussed. Furthermore, this paper will also discuss the effects of process parameters for transesterification in a membrane reactor and the advantages offered by membrane reactors for biodiesel production. This discussion is followed by some limitations faced in membrane technology. Nevertheless, based on the findings presented in this review, it is clear that the membrane reactor has the potential to be a breakthrough technology for the biodiesel industry. Copyright © 2012 Elsevier Inc. All rights reserved.

PROTEUS-SN User Manual

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

Shemon, Emily R.; Smith, Micheal A.; Lee, Changho

2016-02-16

PROTEUS-SN is a three-dimensional, highly scalable, high-fidelity neutron transport code developed at Argonne National Laboratory. The code is applicable to all spectrum reactor transport calculations, particularly those in which a high degree of fidelity is needed either to represent spatial detail or to resolve solution gradients. PROTEUS-SN solves the second order formulation of the transport equation using the continuous Galerkin finite element method in space, the discrete ordinates approximation in angle, and the multigroup approximation in energy. PROTEUS-SN’s parallel methodology permits the efficient decomposition of the problem by both space and angle, permitting large problems to run efficiently on hundredsmore » of thousands of cores. PROTEUS-SN can also be used in serial or on smaller compute clusters (10’s to 100’s of cores) for smaller homogenized problems, although it is generally more computationally expensive than traditional homogenized methodology codes. PROTEUS-SN has been used to model partially homogenized systems, where regions of interest are represented explicitly and other regions are homogenized to reduce the problem size and required computational resources. PROTEUS-SN solves forward and adjoint eigenvalue problems and permits both neutron upscattering and downscattering. An adiabatic kinetics option has recently been included for performing simple time-dependent calculations in addition to standard steady state calculations. PROTEUS-SN handles void and reflective boundary conditions. Multigroup cross sections can be generated externally using the MC2-3 fast reactor multigroup cross section generation code or internally using the cross section application programming interface (API) which can treat the subgroup or resonance table libraries. PROTEUS-SN is written in Fortran 90 and also includes C preprocessor definitions. The code links against the PETSc, METIS, HDF5, and MPICH libraries. It optionally links against the MOAB library and is a part of the SHARP multi-physics suite for coupled multi-physics analysis of nuclear reactors. This user manual describes how to set up a neutron transport simulation with the PROTEUS-SN code. A companion methodology manual describes the theory and algorithms within PROTEUS-SN.« less

Cooling molten salt reactors using "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

2014-08-01

This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

Filamentous bacteria existence in aerobic granular reactors.

PubMed

Figueroa, M; Val del Río, A; Campos, J L; Méndez, R; Mosquera-Corral, A

2015-05-01

Filamentous bacteria are associated to biomass settling problems in wastewater treatment plants. In systems based on aerobic granular biomass they have been proposed to contribute to the initial biomass aggregation process. However, their development on mature aerobic granular systems has not been sufficiently studied. In the present research work, filamentous bacteria were studied for the first time after long-term operation (up to 300 days) of aerobic granular systems. Chloroflexi and Sphaerotilus natans have been observed in a reactor fed with synthetic wastewater. These filamentous bacteria could only come from the inoculated sludge. Thiothrix and Chloroflexi bacteria were observed in aerobic granular biomass treating wastewater from a fish canning industry. Meganema perideroedes was detected in a reactor treating wastewater from a plant processing marine products. As a conclusion, the source of filamentous bacteria in these mature aerobic granular systems fed with industrial effluents was the incoming wastewater.

Decreasing of BOD Concentration on Artificial Domestic Wastewater Using Anaerob Biofilter Reactor Technology

NASA Astrophysics Data System (ADS)

Sumiyati, Sri; Purwanto; Sudarno

2018-02-01

Pollution of domestic wastewater becomes an urban problem. Domestic wastewater contains a variety of pollutants. One of the pollutant parameters in domestic wastewater is BOD. Domestic wastewater which BOD concentrations exceeding the quality standard will be harmful to the environment, particularly the receiving water body. Therefore, before being discharged into the environment, domestic wastewater needs to be processed first. One of the processing that has high efficiency, low cost and easy operation is biofilter technology. The purpose of this research was to analyze the efficiency of BOD concentration reduction in domestic wastewater with anaerobic reactor biofilter using volcanic gravel media. The type of reactor used is an anaerobic biofilter made of glass which volume of 30 liters while the biofilter media is volcanic gravel. In this research the established HRT were 24, 12, 6 and 3 hours. The results showed that the efficiency of BOD concentration reduction in artificial domestic wastewater reached 80%.

Evaluation of performance of select fusion experiments and projected reactors

NASA Technical Reports Server (NTRS)

Miley, G. H.

1978-01-01

The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters.

PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

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

Vondy, D.R.

1981-09-01

This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity.more » The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.« less

Computed tomography of radioactive objects and materials

NASA Astrophysics Data System (ADS)

Sawicka, B. D.; Murphy, R. V.; Tosello, G.; Reynolds, P. W.; Romaniszyn, T.

1990-12-01

Computed tomography (CT) has been performed on a number of radioactive objects and materials. Several unique technical problems are associated with CT of radioactive specimens. These include general safety considerations, techniques to reduce background-radiation effects on CT images and selection criteria for the CT source to permit object penetration and to reveal accurate values of material density. In the present paper, three groups of experiments will be described, for objects with low, medium and high levels of radioactivity. CT studies on radioactive specimens will be presented. They include the following: (1) examination of individual ceramic reactor-fuel (uranium dioxide) pellets, (2) examination of fuel samples from the Three Mile Island reactor, (3) examination of a CANDU (CANada Deuterium Uraniun: registered trademark) nuclear-fuel bundle which underwent a simulated loss-of-coolant accident resulting in high-temperature damage and (4) examination of a PWR nuclear-reactor fuel assembly.

Adaptive Core Simulation Employing Discrete Inverse Theory - Part I: Theory

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

Abdel-Khalik, Hany S.; Turinsky, Paul J.

2005-07-15

Use of adaptive simulation is intended to improve the fidelity and robustness of important core attribute predictions such as core power distribution, thermal margins, and core reactivity. Adaptive simulation utilizes a selected set of past and current reactor measurements of reactor observables, i.e., in-core instrumentation readings, to adapt the simulation in a meaningful way. A meaningful adaption will result in high-fidelity and robust adapted core simulator models. To perform adaption, we propose an inverse theory approach in which the multitudes of input data to core simulators, i.e., reactor physics and thermal-hydraulic data, are to be adjusted to improve agreement withmore » measured observables while keeping core simulator models unadapted. At first glance, devising such adaption for typical core simulators with millions of input and observables data would spawn not only several prohibitive challenges but also numerous disparaging concerns. The challenges include the computational burdens of the sensitivity-type calculations required to construct Jacobian operators for the core simulator models. Also, the computational burdens of the uncertainty-type calculations required to estimate the uncertainty information of core simulator input data present a demanding challenge. The concerns however are mainly related to the reliability of the adjusted input data. The methodologies of adaptive simulation are well established in the literature of data adjustment. We adopt the same general framework for data adjustment; however, we refrain from solving the fundamental adjustment equations in a conventional manner. We demonstrate the use of our so-called Efficient Subspace Methods (ESMs) to overcome the computational and storage burdens associated with the core adaption problem. We illustrate the successful use of ESM-based adaptive techniques for a typical boiling water reactor core simulator adaption problem.« less

SPLASH program for three dimensional fluid dynamics with free surface boundaries

NASA Astrophysics Data System (ADS)

Yamaguchi, A.

1996-05-01

This paper describes a three dimensional computer program SPLASH that solves Navier-Stokes equations based on the Arbitrary Lagrangian Eulerian (ALE) finite element method. SPLASH has been developed for application to the fluid dynamics problems including the moving boundary of a liquid metal cooled Fast Breeder Reactor (FBR). To apply SPLASH code to the free surface behavior analysis, a capillary model using a cubic Spline function has been developed. Several sample problems, e.g., free surface oscillation, vortex shedding development, and capillary tube phenomena, are solved to verify the computer program. In the analyses, the numerical results are in good agreement with the theoretical value or experimental observance. Also SPLASH code has been applied to an analysis of a free surface sloshing experiment coupled with forced circulation flow in a rectangular tank. This is a simplified situation of the flow field in a reactor vessel of the FBR. The computational simulation well predicts the general behavior of the fluid flow inside and the free surface behavior. Analytical capability of the SPLASH code has been verified in this study and the application to more practical problems such as FBR design and safety analysis is under way.

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

Code of Federal Regulations, 2013 CFR

2013-01-01

... radioactive cargo —Function and characteristics of the shipping casks —Radiation hazards —Federal, State and... Contingencies —Accidents —Severe weather conditions —Vehicle breakdown —Communications problems —Radioactive...

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

Code of Federal Regulations, 2014 CFR

2014-01-01

... radioactive cargo —Function and characteristics of the shipping casks —Radiation hazards —Federal, State and... Contingencies —Accidents —Severe weather conditions —Vehicle breakdown —Communications problems —Radioactive...

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

Code of Federal Regulations, 2012 CFR

2012-01-01

... radioactive cargo —Function and characteristics of the shipping casks —Radiation hazards —Federal, State and... Contingencies —Accidents —Severe weather conditions —Vehicle breakdown —Communications problems —Radioactive...

Un cosmologiste oublié: Jean Henri Lambert

NASA Astrophysics Data System (ADS)

Débarbat, Suzanne; Lévy, Jacques

Si les travaux de Kepler ont eu une large influence sure les progrès réalisés en astronomie au cours du 17e siècle, le Siècle de lumières a vu apparaître de nouvelles conceptions. La court vie de J.H. lambert s'inscrit dans le 18e siècle. Il s'agit d'un nom bien connu dans différents domaines (photométrie, projections cartographiques, mathématiques appliquées, etc.); mais il n'est guàre mentionné en cosmologie, alors que Lambert y a fourni une contribution originale offrant quelques suprenantes anticipations...

A Selection of Test Cases for the Validation of Large-Eddy Simulations of Turbulent Flows (Quelques cas d’essai pour la validation de la simulation des gros tourbillons dans les ecoulements turbulents)

DTIC Science & Technology

1998-04-01

they approach the more useful (higher) Reynolds numbers. 8.6 SUMMARY OF COMPLEX FLOWS SQUARE DUCT CMPO00 UDOv 6.5 x 10’i E Yokosawa ei al. 164] pg...Sheets for: Chapter 8. Complex Flows 184 185 CMPOO: Flow in a square duct - Experiments Yokosawa , Fujita, Hirota, & Iwata 1. Description of the flow...These are the experiments of Yokosawa ei al (1989). Air was blown through a flow meter and a settling chamber into a square duct. Measuremsents were

Reconstruction du Flux d'Energie et Recherche de Squarks et Gluinos dans l'Experience D0 (in French)

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

Ridel, Melissa

2002-01-01

Le modèle standard décrit la matière et les interactions fondamentales qui la gouvernent (électromagnétique, faible et forte). L'analyse des données accumulées jusqu'à présent conffrme ces prédictions notamment les mesures de précision effectuées à LEP. Malgré tout, il doit se confronter à quelques dicultés théoriques qui laisseraient penser que le Modèle Standard n'est que la théorie effective d'une autre théorie à plus haute énergie....

Study of Behavior of Some Varieties of Belgian Potatoes Subjected to Gamma Irradiation; ETUDE DU COMPORTEMENT DE QUELQUES VARIETES BELGES DE POMMES DE TERRE SOUMISES A L'IRRADIATION GAMMA

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

Kirchmann, R.; De Proost, M.; Demalsy, P.

1962-07-01

Different varieties of potatoes were irradiated with doses between 5000 and 20000 rads and stored at two different temperatures. Irradiation has a grent influence on the weight loss of the potatoes during storage; the degree of sprout inhibition depends on the variety of the potatoes. The glutathione content and the oxygen consumption of potatoes are influenced by irradiation. The greatest effect of irradiation on the chemical composition concerns the starch; an increase in sugar content is observed. The culinary properties of potatoes are not changed by irradiation. (auth)

Heuristic rules embedded genetic algorithm for in-core fuel management optimization

NASA Astrophysics Data System (ADS)

Alim, Fatih

The objective of this study was to develop a unique methodology and a practical tool for designing loading pattern (LP) and burnable poison (BP) pattern for a given Pressurized Water Reactor (PWR) core. Because of the large number of possible combinations for the fuel assembly (FA) loading in the core, the design of the core configuration is a complex optimization problem. It requires finding an optimal FA arrangement and BP placement in order to achieve maximum cycle length while satisfying the safety constraints. Genetic Algorithms (GA) have been already used to solve this problem for LP optimization for both PWR and Boiling Water Reactor (BWR). The GA, which is a stochastic method works with a group of solutions and uses random variables to make decisions. Based on the theories of evaluation, the GA involves natural selection and reproduction of the individuals in the population for the next generation. The GA works by creating an initial population, evaluating it, and then improving the population by using the evaluation operators. To solve this optimization problem, a LP optimization package, GARCO (Genetic Algorithm Reactor Code Optimization) code is developed in the framework of this thesis. This code is applicable for all types of PWR cores having different geometries and structures with an unlimited number of FA types in the inventory. To reach this goal, an innovative GA is developed by modifying the classical representation of the genotype. To obtain the best result in a shorter time, not only the representation is changed but also the algorithm is changed to use in-core fuel management heuristics rules. The improved GA code was tested to demonstrate and verify the advantages of the new enhancements. The developed methodology is explained in this thesis and preliminary results are shown for the VVER-1000 reactor hexagonal geometry core and the TMI-1 PWR. The improved GA code was tested to verify the advantages of new enhancements. The core physics code used for VVER in this research is Moby-Dick, which was developed to analyze the VVER by SKODA Inc. The SIMULATE-3 code, which is an advanced two-group nodal code, is used to analyze the TMI-1.

Attrition Resistant Iron-Based Fischer-Tropsch Catalysts

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

Jothimurugesan, K.; Goodwin, J.G.; Spivey, J.J.

1997-03-26

The Fischer-Tropsch (F-T) reaction provides a way of converting coal-derived synthesis gas (CO+H{sub 2}) to liquid fuels. Since the reaction is highly exothermic, one of the major problems in control of the reaction is heat removal. Recent work has shown that the use of slurry bubble column reactors (SBCRS) can largely solve this problem. Iron-based (Fe) catalysts are preferred catalysts for F-T when using low CO/H{sub 2} ratio synthesis gases derived from modem coal gasifiers. This is because in addition to reasonable F-T activity, the FT catalysts also possess high water gas shift (WGS) activity. However, a serious problem withmore » the use of Fe catalysts in a SBCR is their tendency to undergo attrition. This can cause fouling/plugging of downstream filters and equipment, making the separation of catalyst from the oil/wax product very difficult if not impossible, and results in a steady loss of catalyst from the reactor. The objectives of this research are to develop a better understanding of the parameters affecting attrition resistance of Fe F-T catalysts suitable for use in SBCRs and to incorporate this understanding into the design of novel Fe catalysts having superior attrition resistance. Catalyst preparations will be based on the use of spray drying and will be scalable using commercially available equipment. The research will employ among other measurements, attrition testing and F-T synthesis, including long duration slurry reactor runs in order to ascertain the degree of success of the various preparations. The goal is to develop an Fe catalyst which can be used in a SBCR having only an internal filter for separation of the catalyst from the liquid product, without sacrificing F-T activity and selectivity.« less

Attrition Resistant Iron-Based Fischer-Tropsch Catalysts.

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

Jothimurugesan, K.; Goodwin, J.S.; Spivey, J.J.

1997-09-22

The Fischer-Tropsch (F-T) reaction provides a way of converting coal-derived synthesis gas (CO and H{sub 2}) to liquid fuels. Since the reaction is highly exothermic, one of the major problems in control of the reaction is heat removal. Recent work has shown that the use of slurry bubble column reactors (SBCRs) can largely solve this problem. Iron-based (Fe) catalysts are preferred catalysts for F-T when using low CO/H{sub 2} ratio synthesis gases derived from modern coal gasifiers. This is because in addition to reasonable F-T activity, the F-T catalysts also possess high water gas shift (WGS) activity. However, a seriousmore » problem with the use of Fe catalysts in a SBCR is their tendency to undergo attrition. This can cause fouling/plugging of downstream filters and equipment, making the separation of catalyst from the oil/wax product very difficult if not impossible, and results in a steady loss of catalyst from the reactor. The objectives of this research are to develop a better understanding of the parameters affecting attrition resistance of Fe F-T catalysts suitable for use in SBCRs and to incorporate this understanding into the design of novel Fe catalysts having superior attrition resistance. Catalyst preparations will be based on the use of spray drying and will be scalable using commercially available equipment. The research will employ among other measurements, attrition testing and F-T synthesis, including long duration slurry reactor runs in order to ascertain the degree of success of the various preparations. The goal is to develop an Fe catalyst which can be used in a SBCR having only an internal filter for separation of the catalyst from the liquid product, without sacrificing F-T activity and selectivity.« less

Analysis and Down Select of Flow Passages for Thermal Hydraulic Testing of a SNAP Derived Reactor

NASA Technical Reports Server (NTRS)

Godfroy, T. J.; Sadasivan, P.; Masterson, S.

2007-01-01

As past of the Vision for Space Exploration, man will return to the moon. To enable safe and productive time on the lunar surface will require adequate power resources. To provide the needed power and to give mission planners all landing site possibilities, including a permanently dark crater, a nuclear reactor provides the most options. Designed to be l00kWt providing approx. 25kWe this power plants would be very effective in delivering dependable, site non-specific power to crews or robotic missions on the lunar surface. An affordable reference reactor based upon the successful SNAP program of the 1960's and early 1970's has been designed by Los Alamos National Laboratory that will meet such a requirement. Considering current funding, environmental, and schedule limitations this lunar surface power reactor will be tested using non-nuclear simulators to simulate the heat from fission reactions. Currently a 25kWe surface power SNAP derivative reactor is in the early process of design and testing with collaboration between Los Alamos National Laboratory, Idaho National Laboratory, Glenn Research Center, Marshall Space Flight Center, and Sandia National Laboratory to ensure that this new design is affordable and can be tested using non-nuclear methods as have proven so effective in the past. This paper will discuss the study and down selection of a flow passage concept for a approx. 25kWe lunar surface power reactor. Several different flow passages designs were evaluated using computational fluid dynamics to determine pressure drop and a structural assessment to consider thermal and stress of the passage walls. The reactor design basis conditions are discussed followed by passage problem setup and results for each concept. A recommendation for passage design is made with rationale for selection.

Direct conversion of nuclear radiation energy

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

Miley, George H.

1970-01-01

This book presents a comprehensive study of methods for converting nuclear radiationi directly without resorting to a heat cycle. The concepts discussed primarily involve direct collection of charged particles released by radioisotopes and by nuclear and thermonuclear reactors. Areas considered include basic energy conversion, charged-particle transport theory, secondary-electron emission, and leakage currents and associated problems. Applications to both nuclear instrumentaion and power sources are discussed. Problems are also included as an aid to the reader or for classroom use.

Teaching Simulation and Modelling at Royal Military College.

ERIC Educational Resources Information Center

Bonin, Hugues W.; Weir, Ronald D.

1984-01-01

Describes a course designed to assist students in writing differential equations to represent chemical processes and to solve these problems on digital computers. Course outline and discussion of computer projects and the simulation and optimization of a continuously stirred tank reactor process are included. (JN)

Report Card on Nuclear Power

ERIC Educational Resources Information Center

Novick, Sheldon

1974-01-01

Problems facing the nuclear power industry include skyrocketing construction costs, technical failures, fuel scarcity, power plant safety, and the disposal of nuclear wastes. Possible solutions include: reductions in nuclear power plant construction, a complete moratorium on new plant construction, the construction of fast breeder reactors and the…

76 FR 80409 - Draft Supplement 2 to Final Environmental Statement Related to the Operation of Watts Bar Nuclear...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-12-23

... image files of the NRC's public documents. If you do not have access to ADAMS or if there are problems.... Stephen J. Campbell, Chief, Watts Bar Special Projects Branch, Division of Operating Reactor Licensing...

Accelerator Driven Nuclear Energy: The Thorium Option

ScienceCinema

Raja, Rajendran

2018-01-05

Conventional nuclear reactors use enriched Uranium as fuel and produce nuclear waste which needs to be stored away for over 10,000 years.   At the current rate of use, existing sources of Uranium will last for 50-100 years.  We describe a solution to the problem that uses particle accelerators to produce fast neutrons that can be used to burn existing nuclear waste and produce energy.  Such systems, initially proposed by Carlo Rubbia and collaborators in the 1990's, are being seriously considered by many countries as a possible solution to the green energy problem.  Accelerator driven reactors operate in a sub-critical regime and, thus, are safer and can obtain energy from plentiful elements such as Thorium-232 and Uranium-238. What is missing is the high intensity (10MW) accelerator that produces 1 GeV protons. We will describe scenarios which if implemented will make such systems a reality.  

High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Ivanov, Aleksandar; Sanchez, Victor; Ivanov, Kostadin

2014-06-01

Monte Carlo methods have been used as reference reactor physics calculation tools worldwide. The advance in computer technology allows the calculation of detailed flux distributions in both space and energy. In most of the cases however, those calculations are done under the assumption of homogeneous material density and temperature distributions. The aim of this work is to develop a consistent methodology for providing realistic three-dimensional thermal-hydraulic distributions by coupling the in-house developed sub-channel code SUBCHANFLOW with the standard Monte-Carlo transport code MCNP. In addition to the innovative technique of on-the fly material definition, a flux-based weight-window technique has been introduced to improve both the magnitude and the distribution of the relative errors. Finally, a coupled code system for the simulation of steady-state reactor physics problems has been developed. Besides the problem of effective feedback data interchange between the codes, the treatment of temperature dependence of the continuous energy nuclear data has been investigated.

Nuclear Power; Past, present and future

NASA Astrophysics Data System (ADS)

Elliott, David

2017-04-01

This book looks at the early history of nuclear power, at what happened next, and at its longer-term prospects. The main question is: can nuclear power overcome the problems that have emerged? It was once touted as the ultimate energy source, freeing mankind from reliance on dirty, expensive fossil energy. Sixty years on, nuclear only supplies around 11.5% of global energy and is being challenged by cheaper energy options. While the costs of renewable sources, like wind and solar, are falling rapidly, nuclear costs have remained stubbornly high. Its development has also been slowed by a range of other problems, including a spate of major accidents, security concerns and the as yet unresolved issue of what to do with the wastes that it produces. In response, a new generation of nuclear reactors is being developed, many of them actually revised versions of the ideas first looked at in the earlier phase. Will this new generation of reactors bring nuclear energy to the forefront of energy production in the future?

Continuous beer fermentation using immobilized yeast cell bioreactor systems.

PubMed

Brányik, Tomás; Vicente, António A; Dostálek, Pavel; Teixeira, José A

2005-01-01

Traditional beer fermentation and maturation processes use open fermentation and lager tanks. Although these vessels had previously been considered indispensable, during the past decades they were in many breweries replaced by large production units (cylindroconical tanks). These have proved to be successful, both providing operating advantages and ensuring the quality of the final beer. Another promising contemporary technology, namely, continuous beer fermentation using immobilized brewing yeast, by contrast, has found only a limited number of industrial applications. Continuous fermentation systems based on immobilized cell technology, albeit initially successful, were condemned to failure for several reasons. These include engineering problems (excess biomass and problems with CO(2) removal, optimization of operating conditions, clogging and channeling of the reactor), unbalanced beer flavor (altered cell physiology, cell aging), and unrealized cost advantages (carrier price, complex and unstable operation). However, recent development in reactor design and understanding of immobilized cell physiology, together with application of novel carrier materials, could provide a new stimulus to both research and application of this promising technology.

ADVANTG An Automated Variance Reduction Parameter Generator, Rev. 1

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

Mosher, Scott W.; Johnson, Seth R.; Bevill, Aaron M.

2015-08-01

The primary objective of ADVANTG is to reduce both the user effort and the computational time required to obtain accurate and precise tally estimates across a broad range of challenging transport applications. ADVANTG has been applied to simulations of real-world radiation shielding, detection, and neutron activation problems. Examples of shielding applications include material damage and dose rate analyses of the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source and High Flux Isotope Reactor (Risner and Blakeman 2013) and the ITER Tokamak (Ibrahim et al. 2011). ADVANTG has been applied to a suite of radiation detection, safeguards, and special nuclear materialmore » movement detection test problems (Shaver et al. 2011). ADVANTG has also been used in the prediction of activation rates within light water reactor facilities (Pantelias and Mosher 2013). In these projects, ADVANTG was demonstrated to significantly increase the tally figure of merit (FOM) relative to an analog MCNP simulation. The ADVANTG-generated parameters were also shown to be more effective than manually generated geometry splitting parameters.« less

Program Helps To Determine Chemical-Reaction Mechanisms

NASA Technical Reports Server (NTRS)

Bittker, D. A.; Radhakrishnan, K.

1995-01-01

General Chemical Kinetics and Sensitivity Analysis (LSENS) computer code developed for use in solving complex, homogeneous, gas-phase, chemical-kinetics problems. Provides for efficient and accurate chemical-kinetics computations and provides for sensitivity analysis for variety of problems, including problems involving honisothermal conditions. Incorporates mathematical models for static system, steady one-dimensional inviscid flow, reaction behind incident shock wave (with boundary-layer correction), and perfectly stirred reactor. Computations of equilibrium properties performed for following assigned states: enthalpy and pressure, temperature and pressure, internal energy and volume, and temperature and volume. Written in FORTRAN 77 with exception of NAMELIST extensions used for input.

The ENABLER - Based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial mass in low Earth orbit and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tommorrow's space propulsion needs.

First results from KamLAND: evidence for reactor antineutrino disappearance.

PubMed

Eguchi, K; Enomoto, S; Furuno, K; Goldman, J; Hanada, H; Ikeda, H; Ikeda, K; Inoue, K; Ishihara, K; Itoh, W; Iwamoto, T; Kawaguchi, T; Kawashima, T; Kinoshita, H; Kishimoto, Y; Koga, M; Koseki, Y; Maeda, T; Mitsui, T; Motoki, M; Nakajima, K; Nakajima, M; Nakajima, T; Ogawa, H; Owada, K; Sakabe, T; Shimizu, I; Shirai, J; Suekane, F; Suzuki, A; Tada, K; Tajima, O; Takayama, T; Tamae, K; Watanabe, H; Busenitz, J; Djurcic, Z; McKinny, K; Mei, D-M; Piepke, A; Yakushev, E; Berger, B E; Chan, Y D; Decowski, M P; Dwyer, D A; Freedman, S J; Fu, Y; Fujikawa, B K; Heeger, K M; Lesko, K T; Luk, K-B; Murayama, H; Nygren, D R; Okada, C E; Poon, A W P; Steiner, H M; Winslow, L A; Horton-Smith, G A; McKeown, R D; Ritter, J; Tipton, B; Vogel, P; Lane, C E; Miletic, T; Gorham, P W; Guillian, G; Learned, J G; Maricic, J; Matsuno, S; Pakvasa, S; Dazeley, S; Hatakeyama, S; Murakami, M; Svoboda, R C; Dieterle, B D; DiMauro, M; Detwiler, J; Gratta, G; Ishii, K; Tolich, N; Uchida, Y; Batygov, M; Bugg, W; Cohn, H; Efremenko, Y; Kamyshkov, Y; Kozlov, A; Nakamura, Y; De Braeckeleer, L; Gould, C R; Karwowski, H J; Markoff, D M; Messimore, J A; Nakamura, K; Rohm, R M; Tornow, W; Young, A R; Wang, Y-F

2003-01-17

KamLAND has measured the flux of nu;(e)'s from distant nuclear reactors. We find fewer nu;(e) events than expected from standard assumptions about nu;(e) propagation at the 99.95% C.L. In a 162 ton.yr exposure the ratio of the observed inverse beta-decay events to the expected number without nu;(e) disappearance is 0.611+/-0.085(stat)+/-0.041(syst) for nu;(e) energies >3.4 MeV. In the context of two-flavor neutrino oscillations with CPT invariance, all solutions to the solar neutrino problem except for the "large mixing angle" region are excluded.

The radiation chemistry of nuclear reactor decontaminating reagents

NASA Astrophysics Data System (ADS)

Sellers, Robin M.

Processes involved in the radiation chemistry of some typical nuclear reactor decontaminating reagents including complexing, reducing and oxidising agents are described. It is concluded that radiation-induced decomposition is only likely to be a problem with dilute formulations, and/or with minor additives such as corrosion inhibitors which are not protected from attack by the other constituents. Addition of a "sacrificial" compound may be necessary to overcome this. The importance of considering loss of function, rather than the decomposition rate of the starting material, is emphasised. Reagents based on low oxidation state metal ions (LOMI) can be regenerated by the radiation field in the presence of formate ion.

Fast Flux Test Facility thermal and pressure transient events during Cycle 11

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

Ahrens, D. M.

1992-03-01

This report documents the thermal and pressure transients experienced by the Reactor Heat Transport System (RHTS) during Cycle 11 which included Cycles 11A, 11B-1, 11B-2 and 11C (i.e. 4 startups and 4 shutdowns). Cycle 11 consisted of a refueling period that began on March 14, 1989 and power operation which began on May 3, 1989 and ended on October 27, 1990. Transients resulted from secondary pump starts/stops while at refueling conditions. The major causes of transients at power were five unplanned reactor scrams from 100% power and problems with Loop 2 DHX Fan Controls During 11A.

Modification of Rhodamine WT tracer tests procedure in activated sludge reactors

NASA Astrophysics Data System (ADS)

Knap, Marta; Balbierz, Piotr

2017-11-01

One of the tracers recommended for use in wastewater treatment plants and natural waters is Rhodamine WT, which is a fluorescent dye, allowing to work at low concentrations, but may be susceptible to sorption to activated sludge flocs and chemical quenching of fluorescence by dissolved water constituents. Additionally raw sewage may contain other natural materials or pollutants exhibiting limited fluorescent properties, which are responsible for background fluorescence interference. This paper presents the proposed modifications to the Rhodamine WT tracer tests procedure in activated sludge reactors, which allow to reduce problems with background fluorescence and tracer loss over time, developed on the basis of conducted laboratory and field experiments.

VERA and VERA-EDU 3.5 Release Notes

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

Sieger, Matt; Salko, Robert K.; Kochunas, Brendan M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems.more » For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.« less

Current drive for stability of thermonuclear plasma reactor

NASA Astrophysics Data System (ADS)

Amicucci, L.; Cardinali, A.; Castaldo, C.; Cesario, R.; Galli, A.; Panaccione, L.; Paoletti, F.; Schettini, G.; Spigler, R.; Tuccillo, A.

2016-01-01

To produce in a thermonuclear fusion reactor based on the tokamak concept a sufficiently high fusion gain together stability necessary for operations represent a major challenge, which depends on the capability of driving non-inductive current in the hydrogen plasma. This request should be satisfied by radio-frequency (RF) power suitable for producing the lower hybrid current drive (LHCD) effect, recently demonstrated successfully occurring also at reactor-graded high plasma densities. An LHCD-based tool should be in principle capable of tailoring the plasma current density in the outer radial half of plasma column, where other methods are much less effective, in order to ensure operations in the presence of unpredictably changes of the plasma pressure profiles. In the presence of too high electron temperatures even at the periphery of the plasma column, as envisaged in DEMO reactor, the penetration of the coupled RF power into the plasma core was believed for long time problematic and, only recently, numerical modelling results based on standard plasma wave theory, have shown that this problem should be solved by using suitable parameter of the antenna power spectrum. We show here further information on the new understanding of the RF power deposition profile dependence on antenna parameters, which supports the conclusion that current can be actively driven over a broad layer of the outer radial half of plasma column, thus enabling current profile control necessary for the stability of a reactor.

Comparison of some characteristics of aerobic granules and sludge flocs from sequencing batch reactors.

PubMed

Li, J; Garny, K; Neu, T; He, M; Lindenblatt, C; Horn, H

2007-01-01

Physical, chemical and biological characteristics were investigated for aerobic granules and sludge flocs from three laboratory-scale sequencing batch reactors (SBRs). One reactor was operated as normal SBR (N-SBR) and two reactors were operated as granular SBRs (G-SBR1 and G-SBR2). G-SBR1 was inoculated with activated sludge and G-SBR2 with granules from the municipal wastewater plant in Garching (Germany). The following major parameters and functions were measured and compared between the three reactors: morphology, settling velocity, specific gravity (SG), sludge volume index (SVI), specific oxygen uptake rate (SOUR), distribution of the volume fraction of extracellular polymeric substances (EPS) and bacteria, organic carbon and nitrogen removal. Compared with sludge flocs, granular sludge had excellent settling properties, good solid-liquid separation, high biomass concentration, simultaneous nitrification and denitrification. Aerobic granular sludge does not have a higher microbial activity and there are some problems including higher effluent suspended solids, lower ratio of VSS/SS and no nitrification at the beginning of cultivation. Measurement with CLSM and additional image analysis showed that EPS glycoconjugates build one main fraction inside the granules. The aerobic granules from G-SBR1 prove to be heavier, smaller and have a higher microbial activity compared with G-SBR2. Furthermore, the granules were more compact, with lower SVI and less filamentous bacteria.

Coupling of kinetic Monte Carlo simulations of surface reactions to transport in a fluid for heterogeneous catalytic reactor modeling.

PubMed

Schaefer, C; Jansen, A P J

2013-02-07

We have developed a method to couple kinetic Monte Carlo simulations of surface reactions at a molecular scale to transport equations at a macroscopic scale. This method is applicable to steady state reactors. We use a finite difference upwinding scheme and a gap-tooth scheme to efficiently use a limited amount of kinetic Monte Carlo simulations. In general the stochastic kinetic Monte Carlo results do not obey mass conservation so that unphysical accumulation of mass could occur in the reactor. We have developed a method to perform mass balance corrections that is based on a stoichiometry matrix and a least-squares problem that is reduced to a non-singular set of linear equations that is applicable to any surface catalyzed reaction. The implementation of these methods is validated by comparing numerical results of a reactor simulation with a unimolecular reaction to an analytical solution. Furthermore, the method is applied to two reaction mechanisms. The first is the ZGB model for CO oxidation in which inevitable poisoning of the catalyst limits the performance of the reactor. The second is a model for the oxidation of NO on a Pt(111) surface, which becomes active due to lateral interaction at high coverages of oxygen. This reaction model is based on ab initio density functional theory calculations from literature.

Research of the rotation effect upon the hydrodynamics and heat and mass transport in a chemical reactor

NASA Astrophysics Data System (ADS)

Gicheva, Natalia I.

2017-11-01

The subject of this research is a chemical reactor for producing tungsten. A physical and mathematical model of fluid motion and heat and mass transfer in a vortex chamber of the chemical reactor under forced and free convection has been described and simulated using two methods. The numerical simulation was carried out in «vortex - stream functions and «velocity - pressure» variables. The velocity field, the mass and the temperature distributions in the reactor were obtained. The influence of a rotation effect upon the hydrodynamics and heat and mass transport was showed. The rotation is important for more uniform distribution of temperature and matter in the vortex chamber. Parametric studies on effects of the Reynolds, Prandtl and Rossbi criteria on the flow characteristics were also performed. Reliability of the calculations was verified by comparing the results obtained by the methods mentioned above. Also, the created model was applied for numerically solving of the classical test problem of the velocity distribution in an annular channel and that of a rotating infinite disk in a stationary liquid. The study findings showed a good agreement with the exact solutions.

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

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

NONE

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

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

NASA Astrophysics Data System (ADS)

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

1988-07-01

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

Anaerobic treatment of distillery spent wash - a study on upflow anaerobic fixed film bioreactor.

PubMed

Acharya, Bhavik K; Mohana, Sarayu; Madamwar, Datta

2008-07-01

Anaerobic digestion of wastewater from a distillery industry having very high COD (1,10,000-1,90,000 mg/L) and BOD (50,000-60,000 mg/L) was studied in a continuously fed, up flow fixed film column reactor using different support materials such as charcoal, coconut coir and nylon fibers under varying hydraulic retention time and organic loading rates. The seed consortium was prepared by enrichment with distillery spent wash in a conventional type reactor having working capacity of 3 L and was used for charging the anaerobic column reactor. Amongst the various support materials studied the reactor having coconut coir could treat distillery spent wash at 8d hydraulic retention time with organic loading rate of 23.25 kg COD m(-3)d(-1) leading to 64% COD reduction with biogas production of 7.2 m3 m(-3)d(-1) having high methane yield without any pretreatment or neutralization of the distillery spent wash. This study indicates fixed film biomethanation of distillery spent wash using coconut coir as the support material appears to be a cost effective and promising technology for mitigating the problems caused by distillery effluent.

Method to Reduce Long-lived Fission Products by Nuclear Transmutations with Fast Spectrum Reactors.

PubMed

Chiba, Satoshi; Wakabayashi, Toshio; Tachi, Yoshiaki; Takaki, Naoyuki; Terashima, Atsunori; Okumura, Shin; Yoshida, Tadashi

2017-10-24

Transmutation of long-lived fission products (LLFPs: 79 Se, 93 Zr, 99 Tc, 107 Pd, 129 I, and 135 Cs) into short-lived or non-radioactive nuclides by fast neutron spectrum reactors without isotope separation has been proposed as a solution to the problem of radioactive wastes disposal. Despite investigation of many methods, such transmutation remains technologically difficult. To establish an effective and efficient transmutation system, we propose a novel neutron moderator material, yttrium deuteride (YD 2 ), to soften the neutron spectrum leaking from the reactor core. Neutron energy spectra and effective half-lives of LLFPs, transmutation rates, and support ratios were evaluated with the continuous-energy Monte Carlo code MVP-II/MVP-BURN and the JENDL-4.0 cross section library. With the YD 2 moderator in the radial blanket and shield regions, effective half-lives drastically decreased from 106 to 102 years and the support ratios reached 1.0 for all six LLFPs. This successful development and implementation of a transmutation system for LLFPs without isotope separation contributes to a the ability of fast spectrum reactors to reduce radioactive waste by consuming their own LLFPs.

Nuclear power: the bargain we can't afford

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

Morgan, R.

1977-01-01

This is a handbook for citizens who wish to raise questions about the costs of atomic energy. It explains, step-by-step, why nuclear reactors have failed to produce low-cost electricity, and it tells citizens how they can use economic arguments to challenge nuclear expansion. Part One, The Costs of Nuclear Energy, contains 7 chapters--The Price of Power (electricity is big business); Mushrooming Capital Costs (nuclear construction costs are skyrocketing); Nuclear Lemons (reactors spend much of their time closed for repairs); The Faulty Fuel Cycle (turning uranium into electricity is not as simple as the utilities say); Hidden Costs (goverment subsidies obscuremore » the true costs of atomic energy); Ratepayer Roulette (nuclear problems translate into higher electric rates); and Alternatives to the Atom (coal-fired power and energy conservation can meet future energy needs more cheaply than nuclear energy). Part Two, Challenging Nuclear Power, contains 3 chapters--Regulators and Reactors (state utility commissions can eliminate the power companies' bias toward nuclear energy); Legislation, Licensing, and Lawsuits (nuclear critics can challenge reactor construction in numerous forums); and Winning the Battle (building an organization is a crucial step in fighting nuclear power). (MCW)« less

Treatment of mountain refuge wastewater by fixed and moving bed biofilm systems.

PubMed

Andreottola, G; Damiani, E; Foladori, P; Nardelli, P; Ragazzi, M

2003-01-01

Tourists visiting mountain refuges in the Alps have increased significantly in the last decade and the number of refuges and huts at high altitude too. In this research the results of an intensive monitoring of a wastewater treatment plant (WWTP) for a tourist mountain refuge located at 2,981 m a.s.l. are described. Two biofilm reactors were adopted: (a) a Moving Bed Biofilm Reactor (MBBR); (b) a submerged Fixed Bed Biofilm Reactor (FBBR). The aims of this research were: (i) the evaluation of the main parameters characterising the processes and involved in the design of the wastewater plants, in order to compare advantages and disadvantages of the two tested alternatives; (ii) the acquisition of an adequate knowledge of the problems connected with the wastewater treatment in alpine refuges. The main results have been: (i) a quick start-up of the biological reactors obtainable thanks to a pre-colonization before the transportation of the plastic carriers to the refuge at the beginning of the tourist season; (ii) low volume and area requirement; (iii) significantly higher removal efficiency compared to other fixed biomass systems, such as trickling filters, but the energy consumption is higher.

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

NASA Astrophysics Data System (ADS)

Butler, Thomas S.

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

Modeling for the optimal biodegradation of toxic wastewater in a discontinuous reactor.

PubMed

Betancur, Manuel J; Moreno-Andrade, Iván; Moreno, Jaime A; Buitrón, Germán; Dochain, Denis

2008-06-01

The degradation of toxic compounds in Sequencing Batch Reactors (SBRs) poses inhibition problems. Time Optimal Control (TOC) methods may be used to avoid such inhibition thus exploiting the maximum capabilities of this class of reactors. Biomass and substrate online measurements, however, are usually unavailable for wastewater applications, so TOC must use only related variables as dissolved oxygen and volume. Although the standard mathematical model to describe the reaction phase of SBRs is good enough for explaining its general behavior in uncontrolled batch mode, better details are needed to model its dynamics when the reactor operates near the maximum degradation rate zone, as when TOC is used. In this paper two improvements to the model are suggested: to include the sensor delay effects and to modify the classical Haldane curve in a piecewise manner. These modifications offer a good solution for a reasonable complexification tradeoff. Additionally, a new way to look at the Haldane K-parameters (micro(o),K(I),K(S)) is described, the S-parameters (micro*,S*,S(m)). These parameters do have a clear physical meaning and, unlike the K-parameters, allow for the statistical treatment to find a single model to fit data from multiple experiments.

REVIEWS OF TOPICAL PROBLEMS: Neutrino oscillations in three- and four-flavor schemes

NASA Astrophysics Data System (ADS)

Akhmedov, Evgenii Kh

2004-02-01

We review some theoretical aspects of neutrino oscillations in the case where more than two neutrino flavors are involved. These include: approximate analytic solutions for 3-flavor (3f) oscillations in matter; matter effects in νμ<-->ντ oscillations; 3f effects in oscillations of solar, atmospheric, reactor, and supernova neutrinos and in accelerator long-baseline experiments; CP and T violation in neutrino oscillations in the vacuum and in matter; the problem of Ue3; and 4f oscillations.

Excore Modeling with VERAShift

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

Pandya, Tara M.; Evans, Thomas M.

It is important to be able to accurately predict the neutron flux outside the immediate reactor core for a variety of safety and material analyses. Monte Carlo radiation transport calculations are required to produce the high fidelity excore responses. Under this milestone VERA (specifically the VERAShift package) has been extended to perform excore calculations by running radiation transport calculations with Shift. This package couples VERA-CS with Shift to perform excore tallies for multiple state points concurrently, with each component capable of parallel execution on independent domains. Specifically, this package performs fluence calculations in the core barrel and vessel, or, performsmore » the requested tallies in any user-defined excore regions. VERAShift takes advantage of the general geometry package in Shift. This gives VERAShift the flexibility to explicitly model features outside the core barrel, including detailed vessel models, detectors, and power plant details. A very limited set of experimental and numerical benchmarks is available for excore simulation comparison. The Consortium for the Advanced Simulation of Light Water Reactors (CASL) has developed a set of excore benchmark problems to include as part of the VERA-CS verification and validation (V&V) problems. The excore capability in VERAShift has been tested on small representative assembly problems, multiassembly problems, and quarter-core problems. VERAView has also been extended to visualize these vessel fluence results from VERAShift. Preliminary vessel fluence results for quarter-core multistate calculations look very promising. Further development is needed to determine the details relevant to excore simulations. Validation of VERA for fluence and excore detectors still needs to be performed against experimental and numerical results.« less

Photoelastic analysis in respect to failure mechanics problems of power plant articles and units

NASA Astrophysics Data System (ADS)

Korikhin, N. V.; Eigenson, S. N.

2009-02-01

The results of strength tests of some critical articles and units of power plants, i.e., a reactor vessel, threaded connection of vessel split, pressure header with straight nipple, turbomachine shaft, and T-weld joint of stator and rotor parts, of turbomachines are presented.

Some Applications of Piece-Wise Smooth Dynamical Systems

NASA Astrophysics Data System (ADS)

Janovská, Drahoslava; Hanus, Tomáš; Biák, Martin

2010-09-01

The Filippov systems theory is applied to selected problems from biology and chemical engineering, namely we explore and simulate Bazykin's ecological model, an ideal closed gas-liquid system including its dimensionless formulation. The last investigated system is a CSTR with an outfall and the CSTR with a reactor volume control.

Production of hydrogen by direct gasification of coal with steam using nuclear heat

NASA Technical Reports Server (NTRS)

1975-01-01

Problems related to: (1) high helium outlet temperature of the reactor, and (2) gas generator design used in hydrogen production are studied. Special attention was given to the use of Oklahoma coal in the gasification process. Plant performance, operation, and environmental considerations are covered.

A framework for modeling and optimizing dynamic systems under uncertainty

DOE PAGES

Nicholson, Bethany; Siirola, John

2017-11-11

Algebraic modeling languages (AMLs) have drastically simplified the implementation of algebraic optimization problems. However, there are still many classes of optimization problems that are not easily represented in most AMLs. These classes of problems are typically reformulated before implementation, which requires significant effort and time from the modeler and obscures the original problem structure or context. In this work we demonstrate how the Pyomo AML can be used to represent complex optimization problems using high-level modeling constructs. We focus on the operation of dynamic systems under uncertainty and demonstrate the combination of Pyomo extensions for dynamic optimization and stochastic programming.more » We use a dynamic semibatch reactor model and a large-scale bubbling fluidized bed adsorber model as test cases.« less

A framework for modeling and optimizing dynamic systems under uncertainty

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

Nicholson, Bethany; Siirola, John

Algebraic modeling languages (AMLs) have drastically simplified the implementation of algebraic optimization problems. However, there are still many classes of optimization problems that are not easily represented in most AMLs. These classes of problems are typically reformulated before implementation, which requires significant effort and time from the modeler and obscures the original problem structure or context. In this work we demonstrate how the Pyomo AML can be used to represent complex optimization problems using high-level modeling constructs. We focus on the operation of dynamic systems under uncertainty and demonstrate the combination of Pyomo extensions for dynamic optimization and stochastic programming.more » We use a dynamic semibatch reactor model and a large-scale bubbling fluidized bed adsorber model as test cases.« less

Study of the Retention of Fission Products by a Few Common Minerals. Application to the Treatment of Medium Activity Effluents; ETUDE DE LA RETENTION DES PRODUITS DE FISSION PAR QUELQUES MINERAUX USUELS. (APPLICATION AUX TRAITEMENTS D'EFFLUENTS DE MOYENNE ACTIVITE SPECIFIQUE)

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

Auchapt, J.M.

1962-01-01

The conditions in which Sr is fixed on calcite (the object of Geneva report P/395-USA-- 1958) are more closely studied and the work is extended to five fission products in the effluerts and to 17 common rocks and minerals. Although this fixation is not suitsble as a method of treating STE effluents (i.e., those from the effluent treatment plant at MIarcoule), the study shows that all the crystals considered are strongly contaminated by simple contact. (auth)

The ENABLER—based on proven NERVA technology

NASA Astrophysics Data System (ADS)

Livingston, Julie M.; Pierce, Bill L.

1991-01-01

The ENABLER reactor for use in a nuclear thermal propulsion engine uses the technology developed in the NERVA/Rover program, updated to incorporate advances in the technology. Using composite fuel, higher power densities per fuel element, improved radiation resistant control components and the advancements in use of carbon-carbon materials; the ENABLER can provide a specific impulse of 925 seconds, an engine thrust to weight (excluding reactor shield) approaching five, an improved initial Mass In Low Earth Orbit (IMLEO) and a consequent reduction in launch costs and logistics problems. This paper describes the 75,000 lbs thrust ENABLER design which is a low cost, low risk approach to meeting tomorrow's space propulsion needs.

Using the surface charge profiler for in-line monitoring of doping concentration in silicon epitaxial wafer manufacturing

NASA Astrophysics Data System (ADS)

Tower, Joshua P.; Kamieniecki, Emil; Nguyen, M. C.; Danel, Adrien

1999-08-01

The Surface Charge Profiler (SCP) has been introduced for monitoring and development of silicon epitaxial processes. The SCP measures the near-surface doping concentration and offers advantages that lead to yield enhancement in several ways. First, non-destructive measurement technology enables in-line process monitoring, eliminating the need to sacrifice production wafers for resistivity measurements. Additionally, the full-wafer mapping capability helps in development of improved epitaxial growth processes and early detection of reactor problems. As examples, we present the use of SCP to study the effects of susceptor degradation in barrel reactors and to study autodoping for development of improved dopant uniformity.

Heat transfer of molten metal layers in severe accidents

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

Wong, Seung Kai; Walton, A.; Yang, Zhilin

1997-12-01

In some scenarios of severe accidents of light water reactors, a layer of molten metal from internal structural components of the pressure vessel is predicted to occur on top of a ceramic core debris in the lower head. The layer transfers the heat generated in the ceramic pool to the side wall of the vessel, causing the latter to melt. This problem has been investigated by Theofanous et al. for the advanced light water reactor AP600 in the context of the accident management strategy of ex-vessel cooling, and the conclusion was drawn that the melting does not seriously compromise themore » integrity of the pressure vessel.« less

DIFFUSE: a FORTRAN program for design computation of tritium transport through thermonuclear reactor components by combined ordinary and thermal diffusion when the principal resistance to diffusion is the bulk metal

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

Pendergrass, J.H.

1977-10-01

Based on the theory developed in an earlier report, a FORTRAN computer program, DIFFUSE, was written. It computes, for design purposes, rates of transport of hydrogen isotopes by temperature-dependent quasi-unidirectional, and quasi-static combined ordinary and thermal diffusion through thin, hot thermonuclear reactor components that can be represented by composites of plane, cylindrical-shell, and spherical-shell elements when the dominant resistance to transfer is that of the bulk metal. The program is described, directions for its use are given, and a listing of the program, together with sample problem results, is presented.

GLSENS: A Generalized Extension of LSENS Including Global Reactions and Added Sensitivity Analysis for the Perfectly Stirred Reactor

NASA Technical Reports Server (NTRS)

Bittker, David A.

1996-01-01

A generalized version of the NASA Lewis general kinetics code, LSENS, is described. The new code allows the use of global reactions as well as molecular processes in a chemical mechanism. The code also incorporates the capability of performing sensitivity analysis calculations for a perfectly stirred reactor rapidly and conveniently at the same time that the main kinetics calculations are being done. The GLSENS code has been extensively tested and has been found to be accurate and efficient. Nine example problems are presented and complete user instructions are given for the new capabilities. This report is to be used in conjunction with the documentation for the original LSENS code.

Source-to-incident-flux relation in a Tokamak blanket module

NASA Astrophysics Data System (ADS)

Imel, G. R.

The next-generation Tokamak experiments, including the Tokamak fusion test reactor (TFTR), will utilize small blanket modules to measure performance parameters such as tritium breeding profiles, power deposition profiles, and neutron flux profiles. Specifically, a neutron calorimeter (simply a neutron moderating blanket module) which permits inferring the incident 14 MeV flux based on measured temperature profiles was proposed for TFTR. The problem of how to relate this total scalar flux to the fusion neutron source is addressed. This relation is necessary since the calorimeter is proposed as a total fusion energy monitor. The methods and assumptions presented was valid for the TFTR Lithium Breeding Module (LBM), as well as other modules on larger Tokamak reactors.

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

Bakosi, Jozsef; Christon, Mark A.; Francois, Marianne M.

Progress is reported on computational capabilities for the grid-to-rod-fretting (GTRF) problem of pressurized water reactors. Numeca's Hexpress/Hybrid mesh generator is demonstrated as an excellent alternative to generating computational meshes for complex flow geometries, such as in GTRF. Mesh assessment is carried out using standard industrial computational fluid dynamics practices. Hydra-TH, a simulation code developed at LANL for reactor thermal-hydraulics, is demonstrated on hybrid meshes, containing different element types. A series of new Hydra-TH calculations has been carried out collecting turbulence statistics. Preliminary results on the newly generated meshes are discussed; full analysis will be documented in the L3 milestone, THM.CFD.P5.05,more » Sept. 2012.« less

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

Hughes, A.C.; Sanders, C.; Tennet, M.G.

""Jason"" reactors are described in which the power level is increased from the original 10 kw to 100 kw. The problems encountered in making this ten- fold increase in power arise not only in connection with the removal of the extra heat produced but also with a number of effects which, although negligible at 10 kw, become significant at 100 kw. These effects are examined and the steps taken, where necessary, to prevent them from becoming troublesome are described. Attention is paid to the safety of the system. A program of work carried out on the Langley ""Jason,"" which throwsmore » considerable light on the behavior of a 100 kw reactor under severe fault conditions, is described here for the first time. (auth)« less

Design of the DEMO Fusion Reactor Following ITER.

PubMed

Garabedian, Paul R; McFadden, Geoffrey B

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Design of the DEMO Fusion Reactor Following ITER

PubMed Central

Garabedian, Paul R.; McFadden, Geoffrey B.

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

Separations in the STATS report

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

Choppin, G.R.

1996-12-31

The Separations Technology and Transmutation Systems (STATS) Committee formed a Subcommittee on Separations. This subcommittee was charged with evaluating the separations proposed for the several reactor and accelerator transmutation systems. It was also asked to review the processing options for the safe management of high-level waste generated by the defense programs, in particular, the special problems involved in dealing with the waste at the U.S. Department of Energy (DOE) facility in Hanford, Washington. Based on the evaluations from the Subcommittee on Separations, the STATS Committee concluded that for the reactor transmutation programs, aqueous separations involving a combination of PUREX andmore » TRUEX solvent extraction processes could be used. However, additional research and development (R&D) would be required before full plant-scale use of the TRUEX technology could be employed. Alternate separations technology for the reactor transmutation program involves pyroprocessing. This process would require a significant amount of R&D before its full-scale application can be evaluated.« less

A Numerical Model for Trickle Bed Reactors

NASA Astrophysics Data System (ADS)

Propp, Richard M.; Colella, Phillip; Crutchfield, William Y.; Day, Marcus S.

2000-12-01

Trickle bed reactors are governed by equations of flow in porous media such as Darcy's law and the conservation of mass. Our numerical method for solving these equations is based on a total-velocity splitting, sequential formulation which leads to an implicit pressure equation and a semi-implicit mass conservation equation. We use high-resolution finite-difference methods to discretize these equations. Our solution scheme extends previous work in modeling porous media flows in two ways. First, we incorporate physical effects due to capillary pressure, a nonlinear inlet boundary condition, spatial porosity variations, and inertial effects on phase mobilities. In particular, capillary forces introduce a parabolic component into the recast evolution equation, and the inertial effects give rise to hyperbolic nonconvexity. Second, we introduce a modification of the slope-limiting algorithm to prevent our numerical method from producing spurious shocks. We present a numerical algorithm for accommodating these difficulties, show the algorithm is second-order accurate, and demonstrate its performance on a number of simplified problems relevant to trickle bed reactor modeling.

Antifoaming effect of chemical compounds in manure biogas reactors.

PubMed

Kougias, P G; Tsapekos, P; Boe, K; Angelidaki, I

2013-10-15

A precise and efficient antifoaming control strategy in bioprocesses is a challenging task as foaming is a very complex phenomenon. Nevertheless, foam control is necessary, as foam is a major operational problem in biogas reactors. In the present study, the effect of 14 chemical compounds on foam reduction was evaluated at concentration of 0.05%, 0.1% and 0.5% v/v(sample), in raw and digested manure. Moreover, two antifoam injection methods were compared for foam reduction efficiency. Natural oils (rapeseed and sunflower oil), fatty acids (oleic, octanoic and derivative of natural fatty acids), siloxanes (polydimethylsiloxane) and ester (tributylphosphate) were found to be the most efficient compounds to suppress foam. The efficiency of antifoamers was dependant on their physicochemical properties and greatly correlated to their chemical characteristics for dissolving foam. The antifoamers were more efficient in reducing foam when added directly into the liquid phase rather than added in the headspace of the reactor. Copyright © 2013 Elsevier Ltd. All rights reserved.

Contributions Regarding the Aircraft Nuclear Propulsion

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

Mitrica, Bogdan; Petre, Marian; Dima, Mihai Octavian

2010-01-21

The possibility to use a nuclear reactor for airplanes propulsion was investigated taking in to account 2 possible solutions: the direct cycle (where the fluid pass through the reactor's core) and the indirect cycle (where the fluid is passing through a heat exchanger). Taking in to account the radioprotection problems, the only realistic solution seems to be the indirect cycle, where the energy transfer should be performed by a heat exchanger that must work at very high speed of the fluid. The heat exchanger will replace the classical burning room. We had performed a more precise theoretical study for themore » nuclear jet engine regarding the performances of the nuclear reactor, of the heat exchanger and of the jet engine. It was taken in to account that in the moment when the burning room is replaced by a heat exchanger, a new model for gasodynamic process from the engine must be performed. Studies regarding the high flow speed heat transfer were performed.« less

Fusion reactor blanket/shield design study

NASA Astrophysics Data System (ADS)

Smith, D. L.; Clemmer, R. G.; Harkness, S. D.; Jung, J.; Krazinski, J. L.; Mattas, R. F.; Stevens, H. C.; Youngdahl, C. K.; Trachsel, C.; Bowers, D.

1979-07-01

A joint study of Tokamak reactor first wall/blanket/shield technology was conducted to identify key technological limitations for various tritium breeding blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium breeding blanket concepts were evaluated according to the proposed coolant. The effort concentrated on evaluation of lithium and water cooled blanket designs and helium and molten salt cooled designs. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a Tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented.

Contributions to nuclear safety and radiation technologies in Ukraine by the Science and Technology Center in Ukraine (STCU)

NASA Astrophysics Data System (ADS)

Taranenko, L.; Janouch, F.; Owsiacki, L.

2001-06-01

This paper presents Science and Technology Center in Ukraine (STCU) activities devoted to furthering nuclear and radiation safety, which is a prioritized STCU area. The STCU, an intergovernmental organization with the principle objective of non-proliferation, administers financial support from the USA, Canada, and the EU to Ukrainian projects in various scientific and technological areas; coordinates projects; and promotes the integration of Ukrainian scientists into the international scientific community, including involving western collaborators. The paper focuses on STCU's largest project to date "Program Supporting Y2K Readiness at Ukrainian NPPs" initiated in April 1999 and designed to address possible Y2K readiness problems at 14 Ukrainian nuclear reactors. Other presented projects demonstrate a wide diversity of supported directions in the fields of nuclear and radiation safety, including reactor material improvement ("Improved Zirconium-Based Elements for Nuclear Reactors"), information technologies for nuclear industries ("Ukrainian Nuclear Data Bank in Slavutich"), and radiation health science ("Diagnostics and Treatment of Radiation-Induced Injuries of Human Biopolymers").

Exposure calculation code module for reactor core analysis: BURNER

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

Vondy, D.R.; Cunningham, G.W.

1979-02-01

The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also providesmore » user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.« less

The feasibility of trace element supplementation for stable operation of wheat stillage-fed biogas tank reactors.

PubMed

Gustavsson, J; Svensson, B H; Karlsson, A

2011-01-01

The aim of this study was to investigate the effect of trace element supplementation on operation of wheat stillage-fed biogas tank reactors. The stillage used was a residue from bio-ethanol production, containing high levels of sulfate. In biogas production, high sulfate content has been associated with poor process stability in terms of low methane production and accumulation of process intermediates. However, the results of the present study show that this problem can be overcome by trace element supplementations. Four lab-scale wheat stillage-fed biogas tank reactors were operated for 345 days at a hydraulic retention time of 20 days (37 degrees C). It was concluded that daily supplementation with Co (0.5 mg L(-1)), Ni (0.2 mg L(-1)) and Fe (0.5 g L(-1)) were required for maintaining process stability at the organic loading rate of 4.0 g volatile solids L(-1) day(-1).

Enhanced methane generation during theromophilic co-digestion of confectionary waste and grease-trap fats and oils with municipal wastewater sludge.

PubMed

Gough, Heidi L; Nelsen, Diane; Muller, Christopher; Ferguson, John

2013-02-01

Recent interest in carbon-neutral biofuels has revived interest in co-digestion for methane generation. At wastewater treatment facilities, organic wastes may be co-digested with sludge using established anaerobic digesters. However, changes to organic loadings may induce digester instability, particularly for thermophilic digesters. To examine this problem, thermophilic (55 degrees C) co-digestion was studied for two food-industry wastes in semi-continuous laboratory digesters; in addition, the wastes' biochemical methane potentials were tested. Wastes with high chemical oxygen demand (COD) content were selected as feedstocks allowing increased input of potential energy to reactors without substantially altering volumetric loadings. Methane generation increased while reactor pH and volatile solids remained stable. Lag periods observed prior to methane stimulation suggested that acclimation of the microbial community may be critical to performance during co-digestion. Chemical oxygen demand mass balances in the experimental and control reactors indicated that all of the food industry waste COD was converted to methane.

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

Rao, Nageswara S.; Ramirez Aviles, Camila A.

We consider the problem of inferring the operational status of a reactor facility using measurements from a radiation sensor network deployed around the facility’s ventilation off-gas stack. The intensity of stack emissions decays with distance, and the sensor counts or measurements are inherently random with parameters determined by the intensity at the sensor’s location. We utilize the measurements to estimate the intensity at the stack, and use it in a one-sided Sequential Probability Ratio Test (SPRT) to infer on/off status of the reactor. We demonstrate the superior performance of this method over conventional majority fusers and individual sensors using (i)more » test measurements from a network of 21 NaI detectors, and (ii) effluence measurements collected at the stack of a reactor facility. We also analytically establish the superior detection performance of the network over individual sensors with fixed and adaptive thresholds by utilizing the Poisson distribution of the counts. We quantify the performance improvements of the network detection over individual sensors using the packing number of the intensity space.« less

Buoyant Filter Bio-Reactor (BFBR)--a novel anaerobic wastewater treatment unit.

PubMed

Panicker, Soosan J; Philipose, M C; Haridas, Ajit

2008-01-01

The Buoyant Filter Bio-Reactor (BFBR) is a novel and very efficient method for the treatment of complex wastewater. Sewage is a complex wastewater containing insoluble COD contributed by fat and proteins. The fat and proteins present in the domestic sewage cause operational problems and underperformance in the Upflow Anaerobic Sludge Blanket Reactor, used now for treating sewage anaerobically. The biogas yield from the BFBR is 0.36 m3/kg COD reduced and the methane content was about 70-80%. Production of methane by anaerobic digestion of organic waste had the benefit of lower energy costs for treatment and is thus environmentally beneficial to the society by providing a clean fuel from renewable feed stocks. The BFBR achieved a COD removal efficiency of 80-90% for an organic loading rate of 4.5 kg/m3/d at a hydraulic retention time of 3.25 hours. The effluent COD was less than 100 mg/l, thus saving on secondary treatment cost. No pretreatment like sedimentation was required for the influent to the BFBR. The BFBR can produce low turbidity effluent as in the activated sludge process (ASP). The land area required for the BFBR treatment plant is less when compared to ASP plant. Hence the problem of scarcity of land for the treatment plant is reduced. The total expenditure for erecting the unit was less than 50% as that of conventional ASP for the same COD removal efficiency including land cost. IWA Publishing 2008.

A Jacobian-Free Newton Krylov Method for Mortar-Discretized Thermomechanical Contact Problems

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

Glen Hansen

2011-07-01

Multibody contact problems are common within the field of multiphysics simulation. Applications involving thermomechanical contact scenarios are also quite prevalent. Such problems can be challenging to solve due to the likelihood of thermal expansion affecting contact geometry which, in turn, can change the thermal behavior of the components being analyzed. This paper explores a simple model of a light water reactor nuclear reactor fuel rod, which consists of cylindrical pellets of uranium dioxide (UO2) fuel sealed within a Zircalloy cladding tube. The tube is initially filled with helium gas, which fills the gap between the pellets and cladding tube. Themore » accurate modeling of heat transfer across the gap between fuel pellets and the protective cladding is essential to understanding fuel performance, including cladding stress and behavior under irradiated conditions, which are factors that affect the lifetime of the fuel. The thermomechanical contact approach developed here is based on the mortar finite element method, where Lagrange multipliers are used to enforce weak continuity constraints at participating interfaces. In this formulation, the heat equation couples to linear mechanics through a thermal expansion term. Lagrange multipliers are used to formulate the continuity constraints for both heat flux and interface traction at contact interfaces. The resulting system of nonlinear algebraic equations are cast in residual form for solution of the transient problem. A Jacobian-free Newton Krylov method is used to provide for fully-coupled solution of the coupled thermal contact and heat equations.« less

Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

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

Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.

Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent researchmore » progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.« less

Microstructural evolution in fast-neutron-irradiated austenitic stainless steels

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

Stoller, R.E.

1987-12-01

The present work has focused on the specific problem of fast-neutron-induced radiation damage to austenitic stainless steels. These steels are used as structural materials in current fast fission reactors and are proposed for use in future fusion reactors. Two primary components of the radiation damage are atomic displacements (in units of displacements per atom, or dpa) and the generation of helium by nuclear transmutation reactions. The radiation environment can be characterized by the ratio of helium to displacement production, the so-called He/dpa ratio. Radiation damage is evidenced microscopically by a complex microstructural evolution and macroscopically by density changes and alteredmore » mechanical properties. The purpose of this work was to provide additional understanding about mechanisms that determine microstructural evolution in current fast reactor environments and to identify the sensitivity of this evolution to changes in the He/dpa ratio. This latter sensitivity is of interest because the He/dpa ratio in a fusion reactor first wall will be about 30 times that in fast reactor fuel cladding. The approach followed in the present work was to use a combination of theoretical and experimental analysis. The experimental component of the work primarily involved the examination by transmission electron microscopy of specimens of a model austenitic alloy that had been irradiated in the Oak Ridge Research Reactor. A major aspect of the theoretical work was the development of a comprehensive model of microstructural evolution. This included explicit models for the evolution of the major extended defects observed in neutron irradiated steels: cavities, Frank faulted loops and the dislocation network. 340 refs., 95 figs., 18 tabs.« less

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

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

NASA Astrophysics Data System (ADS)

Raj, Baldev; Rao, K. Bhanu Sankara

2009-04-01

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

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

Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

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

Hu, Rui

2017-09-03

Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developedmore » at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.« less

Microwave Plasma Hydrogen Recovery System

NASA Technical Reports Server (NTRS)

Atwater, James; Wheeler, Richard, Jr.; Dahl, Roger; Hadley, Neal

2010-01-01

A microwave plasma reactor was developed for the recovery of hydrogen contained within waste methane produced by Carbon Dioxide Reduction Assembly (CRA), which reclaims oxygen from CO2. Since half of the H2 reductant used by the CRA is lost as CH4, the ability to reclaim this valuable resource will simplify supply logistics for longterm manned missions. Microwave plasmas provide an extreme thermal environment within a very small and precisely controlled region of space, resulting in very high energy densities at low overall power, and thus can drive high-temperature reactions using equipment that is smaller, lighter, and less power-consuming than traditional fixed-bed and fluidized-bed catalytic reactors. The high energy density provides an economical means to conduct endothermic reactions that become thermodynamically favorable only at very high temperatures. Microwave plasma methods were developed for the effective recovery of H2 using two primary reaction schemes: (1) methane pyrolysis to H2 and solid-phase carbon, and (2) methane oligomerization to H2 and acetylene. While the carbon problem is substantially reduced using plasma methods, it is not completely eliminated. For this reason, advanced methods were developed to promote CH4 oligomerization, which recovers a maximum of 75 percent of the H2 content of methane in a single reactor pass, and virtually eliminates the carbon problem. These methods were embodied in a prototype H2 recovery system capable of sustained high-efficiency operation. NASA can incorporate the innovation into flight hardware systems for deployment in support of future long-duration exploration objectives such as a Space Station retrofit, Lunar outpost, Mars transit, or Mars base. The primary application will be for the recovery of hydrogen lost in the Sabatier process for CO2 reduction to produce water in Exploration Life Support systems. Secondarily, this process may also be used in conjunction with a Sabatier reactor employed to stockpile life-support oxygen as well as propellant and fuel production from Martian atmospheric CO2

Current problems in applied mathematics and mathematical modeling

NASA Astrophysics Data System (ADS)

Alekseev, A. S.

Papers are presented on mathematical modeling noting applications to such fields as geophysics, chemistry, atmospheric optics, and immunology. Attention is also given to models of ocean current fluxes, atmospheric and marine interactions, and atmospheric pollution. The articles include studies of catalytic reactors, models of global climate phenomena, and computer-assisted atmospheric models.

The Role of Nuclear Power in Achieving the World We Want

ERIC Educational Resources Information Center

Driscoll, M. J.

1970-01-01

Supports the development of nuclear power plants and considers some problems and possible solutions: future power needs, power costs, thermal pollution, radionuclide discharge. Describes advantages and applications of dual purpose power plants for purifying water, producing phosphorus and ammonia, and serving as fast breeder reactors for Pu 239.…

Fundamental Research in Engineering Education. Student Learning in Industrially Situated Virtual Laboratories

ERIC Educational Resources Information Center

Koretsky, Milo D.; Kelly, Christine; Gummer, Edith

2011-01-01

The instructional design and the corresponding research on student learning of two virtual laboratories that provide an engineering task situated in an industrial context are described. In this problem-based learning environment, data are generated dynamically based on each student team's distinct choices of reactor parameters and measurements.…

78 FR 7818 - Duane Arnold Energy Center; Application for Amendment to Facility Operating License

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-04

... methods: Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013... search, select ``ADAMS Public Documents'' and then select ``Begin Web- based ADAMS Search.'' For problems... INFORMATION CONTACT: Karl D. Feintuch, Project Manager, Office of Nuclear Reactor Regulation, U.S. Nuclear...

77 FR 67837 - Callaway Plant, Unit 1; Application for Amendment to Facility Operating License

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-14

... methods: Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012... search, select ``ADAMS Public Documents'' and then select ``Begin Web- based ADAMS Search.'' For problems... INFORMATION CONTACT: Carl F. Lyon, Project Manager, Office of Nuclear Reactor Regulation, U.S. Nuclear...

Calculating Pi Using the Monte Carlo Method

ERIC Educational Resources Information Center

Williamson, Timothy

2013-01-01

During the summer of 2012, I had the opportunity to participate in a research experience for teachers at the center for sustainable energy at Notre Dame University (RET @ cSEND) working with Professor John LoSecco on the problem of using antineutrino detection to accurately determine the fuel makeup and operating power of nuclear reactors. During…

VERA 3.6 Release Notes

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

Williamson, Richard L.; Kochunas, Brendan; Adams, Brian M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms.

Tungsten - Yttrium Based Nuclear Structural Materials

NASA Astrophysics Data System (ADS)

Ramana, Chintalapalle; Chessa, Jack; Martinenz, Gustavo

2013-04-01

The challenging problem currently facing the nuclear science community in this 21st century is design and development of novel structural materials, which will have an impact on the next-generation nuclear reactors. The materials available at present include reduced activation ferritic/martensitic steels, dispersion strengthened reduced activation ferritic steels, and vanadium- or tungsten-based alloys. These materials exhibit one or more specific problems, which are either intrinsic or caused by reactors. This work is focussed towards tungsten-yttrium (W-Y) based alloys and oxide ceramics, which can be utilized in nuclear applications. The goal is to derive a fundamental scientific understanding of W-Y-based materials. In collaboration with University of Califonia -- Davis, the project is designated to demonstrate the W-Y based alloys, ceramics and composites with enhanced physical, mechanical, thermo-chemical properties and higher radiation resistance. Efforts are focussed on understanding the microstructure, manipulating materials behavior under charged-particle and neutron irradiation, and create a knowledge database of defects, elemental diffusion/segregation, and defect trapping along grain boundaries and interfaces. Preliminary results will be discussed.

The search of the best mode of the reserve power supply consumption during the nuclear reactor’s emergency shutdown procedures in case of force majeure circumstances

NASA Astrophysics Data System (ADS)

Zagrebaev, A. M.; Trifonenkov, A. V.

2017-01-01

This article deals with the problem of the control mode choice for a power supply system in case of force majeure circumstances. It is not known precisely, when a force majeure incident occurs, but the threatened period is given, when the incident is expected. It is supposed, that force majeure circumstances force nuclear reactor shutdown at the moment of threat coming. In this article the power supply system is considered, which consists of a nuclear reactor and a reserve power supply, for example, a hydroelectric pumped storage power station. The reserve power supply has limited capacity and it doesn’t undergo the threatened incident. The problem of the search of the best reserve supply time-distribution in case of force majeure circumstances is stated. The search is performed according to minimization of power loss and damage to the infrastructure. The software has been developed, which performs automatic numerical search of the approximate optimal control modes for the reserve power supply.

A Perspective on Coupled Multiscale Simulation and Validation in Nuclear Materials

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

M. P. Short; D. Gaston; C. R. Stanek

2014-01-01

The field of nuclear materials encompasses numerous opportunities to address and ultimately solve longstanding industrial problems by improving the fundamental understanding of materials through the integration of experiments with multiscale modeling and high-performance simulation. A particularly noteworthy example is an ongoing study of axial power distortions in a nuclear reactor induced by corrosion deposits, known as CRUD (Chalk River unidentified deposits). We describe how progress is being made toward achieving scientific advances and technological solutions on two fronts. Specifically, the study of thermal conductivity of CRUD phases has augmented missing data as well as revealed new mechanisms. Additionally, the developmentmore » of a multiscale simulation framework shows potential for the validation of a new capability to predict the power distribution of a reactor, in effect direct evidence of technological impact. The material- and system-level challenges identified in the study of CRUD are similar to other well-known vexing problems in nuclear materials, such as irradiation accelerated corrosion, stress corrosion cracking, and void swelling; they all involve connecting materials science fundamentals at the atomistic- and mesoscales to technology challenges at the macroscale.« less

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

NASA Astrophysics Data System (ADS)

Shadid, J. N.; Smith, T. M.; Cyr, E. C.; Wildey, T. M.; Pawlowski, R. P.

2016-09-01

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts to apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier-Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.

A Generalized Perturbation Theory Solver In Rattlesnake Based On PETSc With Application To TREAT Steady State Uncertainty Quantification

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

Schunert, Sebastian; Wang, Congjian; Wang, Yaqi

Rattlesnake and MAMMOTH are the designated TREAT analysis tools currently being developed at the Idaho National Laboratory. Concurrent with development of the multi-physics, multi-scale capabilities, sensitivity analysis and uncertainty quantification (SA/UQ) capabilities are required for predicitive modeling of the TREAT reactor. For steady-state SA/UQ, that is essential for setting initial conditions for the transients, generalized perturbation theory (GPT) will be used. This work describes the implementation of a PETSc based solver for the generalized adjoint equations that constitute a inhomogeneous, rank deficient problem. The standard approach is to use an outer iteration strategy with repeated removal of the fundamental modemore » contamination. The described GPT algorithm directly solves the GPT equations without the need of an outer iteration procedure by using Krylov subspaces that are orthogonal to the operator’s nullspace. Three test problems are solved and provide sufficient verification for the Rattlesnake’s GPT capability. We conclude with a preliminary example evaluating the impact of the Boron distribution in the TREAT reactor using perturbation theory.« less

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

James K. Neathery; Gary Jacobs; Burtron H. Davis

In this reporting period, a fundamental filtration study was started to investigate the separation of Fischer-Tropsch Synthesis (FTS) liquids from iron-based catalyst particles. Slurry-phase FTS in slurry bubble column reactor systems is the preferred mode of production since the reaction is highly exothermic. Consequently, heavy wax products must be separated from catalyst particles before being removed from the reactor system. Achieving an efficient wax product separation from iron-based catalysts is one of the most challenging technical problems associated with slurry-phase FTS. The separation problem is further compounded by catalyst particle attrition and the formation of ultra-fine iron carbide and/or carbonmore » particles. Existing pilot-scale equipment was modified to include a filtration test apparatus. After undergoing an extensive plant shakedown period, filtration tests with cross-flow filter modules using simulant FTS wax slurry were conducted. The focus of these early tests was to find adequate mixtures of polyethylene wax to simulate FTS wax. Catalyst particle size analysis techniques were also developed. Initial analyses of the slurry and filter permeate particles will be used by the research team to design improved filter media and cleaning strategies.« less

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

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

Shadid, J.N., E-mail: jnshadi@sandia.gov; Department of Mathematics and Statistics, University of New Mexico; Smith, T.M.

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. In this respect the understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In this study we report on initial efforts tomore » apply integrated adjoint-based computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. Initial results are presented that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.« less

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

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

Shadid, J. N.; Smith, T. M.; Cyr, E. C.

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. The understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In our study we report on initial efforts to apply integrated adjoint-basedmore » computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. We present the initial results that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.« less

Stabilized FE simulation of prototype thermal-hydraulics problems with integrated adjoint-based capabilities

DOE PAGES

Shadid, J. N.; Smith, T. M.; Cyr, E. C.; ...

2016-05-20

A critical aspect of applying modern computational solution methods to complex multiphysics systems of relevance to nuclear reactor modeling, is the assessment of the predictive capability of specific proposed mathematical models. The understanding of numerical error, the sensitivity of the solution to parameters associated with input data, boundary condition uncertainty, and mathematical models is critical. Additionally, the ability to evaluate and or approximate the model efficiently, to allow development of a reasonable level of statistical diagnostics of the mathematical model and the physical system, is of central importance. In our study we report on initial efforts to apply integrated adjoint-basedmore » computational analysis and automatic differentiation tools to begin to address these issues. The study is carried out in the context of a Reynolds averaged Navier–Stokes approximation to turbulent fluid flow and heat transfer using a particular spatial discretization based on implicit fully-coupled stabilized FE methods. We present the initial results that show the promise of these computational techniques in the context of nuclear reactor relevant prototype thermal-hydraulics problems.« less

Initial experimental evaluation of crud-resistant materials for light water reactors

NASA Astrophysics Data System (ADS)

Dumnernchanvanit, I.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Carlson, M. B.; Hussey, D.; Short, M. P.

2018-01-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

Implementation of ALARA at the design stage of Nuclear Power Plants

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

Brissaud, A.; Ridoux, P.

1995-03-01

In the 1970s, Electricite de France (EdF) had limited knowledge and experience of pressurized water reactors (PWRs). Electricity generation by nuclear units was oriented towards gas-graphite reactors, even though EdF had a share in the PWR unit of CHOOZ A-1 (250 MWe, later upgraded to 320 MWe). Some facts about the origin of doses in that king of reactor were known to the research and development (R&D) support staff of EdF, which mainly comprises the French Atomic Commission (CEA), but only a few of EdF`s engineers were aware of these facts. One has to bear in mind that CHOOZ A-1more » only went critical in April 1967 and was officially connected to the grid in May 1970 after some important problems had been solved. Meanwhile, the nuclear program was launched at full speed, beginning with the order for FESSENHEIM 1 in 1970, FESSENHEIM 2 and BUGEY 2 and 3 in 1971. TIHANGE 1, in which EdF had a share, went on-line in September 1975. Also, supposing that EdF had had such knowledge and experience, it is quite evident that it would have been very difficult to modify the lay-out inside the reactor building.« less

Inductive System for Reliable Magnesium Level Detection in a Titanium Reduction Reactor

NASA Astrophysics Data System (ADS)

Krauter, Nico; Eckert, Sven; Gundrum, Thomas; Stefani, Frank; Wondrak, Thomas; Frick, Peter; Khalilov, Ruslan; Teimurazov, Andrei

2018-05-01

The determination of the Magnesium level in a Titanium reduction retort by inductive methods is often hampered by the formation of Titanium sponge rings which disturb the propagation of electromagnetic signals between excitation and receiver coils. We present a new method for the reliable identification of the Magnesium level which explicitly takes into account the presence of sponge rings with unknown geometry and conductivity. The inverse problem is solved by a look-up-table method, based on the solution of the inductive forward problems for several tens of thousands parameter combinations.

Large-eddy simulations of turbulent flow for grid-to-rod fretting in nuclear reactors

DOE PAGES

Bakosi, J.; Christon, M. A.; Lowrie, R. B.; ...

2013-07-12

The grid-to-rod fretting (GTRF) problem in pressurized water reactors is a flow-induced vibration problem that results in wear and failure of the fuel rods in nuclear assemblies. In order to understand the fluid dynamics of GTRF and to build an archival database of turbulence statistics for various configurations, implicit large-eddy simulations of time-dependent single-phase turbulent flow have been performed in 3 × 3 and 5 × 5 rod bundles with a single grid spacer. To assess the computational mesh and resolution requirements, a method for quantitative assessment of unstructured meshes with no-slip walls is described. The calculations have been carriedmore » out using Hydra-TH, a thermal-hydraulics code developed at Los Alamos for the Consortium for Advanced Simulation of Light water reactors, a United States Department of Energy Innovation Hub. Hydra-TH uses a second-order implicit incremental projection method to solve the singlephase incompressible Navier-Stokes equations. The simulations explicitly resolve the large scale motions of the turbulent flow field using first principles and rely on a monotonicity-preserving numerical technique to represent the unresolved scales. Each series of simulations for the 3 × 3 and 5 × 5 rod-bundle geometries is an analysis of the flow field statistics combined with a mesh-refinement study and validation with available experimental data. Our primary focus is the time history and statistics of the forces loading the fuel rods. These hydrodynamic forces are believed to be the key player resulting in rod vibration and GTRF wear, one of the leading causes for leaking nuclear fuel which costs power utilities millions of dollars in preventive measures. As a result, we demonstrate that implicit large-eddy simulation of rod-bundle flows is a viable way to calculate the excitation forces for the GTRF problem.« less

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.

Impact of Nuclear Data Uncertainties on Advanced Fuel Cycles and their Irradiated Fuel - a Comparison between Libraries

NASA Astrophysics Data System (ADS)

Díez, C. J.; Cabellos, O.; Martínez, J. S.

2014-04-01

The uncertainties on the isotopic composition throughout the burnup due to the nuclear data uncertainties are analysed. The different sources of uncertainties: decay data, fission yield and cross sections; are propagated individually, and their effect assessed. Two applications are studied: EFIT (an ADS-like reactor) and ESFR (Sodium Fast Reactor). The impact of the uncertainties on cross sections provided by the EAF-2010, SCALE6.1 and COMMARA-2.0 libraries are compared. These Uncertainty Quantification (UQ) studies have been carried out with a Monte Carlo sampling approach implemented in the depletion/activation code ACAB. Such implementation has been improved to overcome depletion/activation problems with variations of the neutron spectrum.

Low cost solar array project 1: Silicon material

NASA Technical Reports Server (NTRS)

Jewett, D. N.; Bates, H. E.; Hill, D. M.

1980-01-01

The low cost production of silicon by deposition of silicon from a hydrogen/chlorosilane mixture is described. Reactor design, reaction vessel support systems (physical support, power control and heaters, and temperature monitoring systems) and operation of the system are reviewed. Testing of four silicon deposition reactors is described, and test data and consequently derived data are given. An 18% conversion of trichlorosilane to silicon was achieved, but average conversion rates were lower than predicted due to incomplete removal of byproduct gases for recycling and silicon oxide/silicon polymer plugging of the gas outlet. Increasing the number of baffles inside the reaction vessel improved the conversion rate. Plans for further design and process improvements to correct the problems encountered are outlined.

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

NASA Astrophysics Data System (ADS)

Durham, J. M.; Poulson, D.; Bacon, J.; Chichester, D. L.; Guardincerri, E.; Morris, C. L.; Plaud-Ramos, K.; Schwendiman, W.; Tolman, J. D.; Winston, P.

2018-04-01

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. Here we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. This application of technology and methods commonly used in high-energy particle physics provides a potential solution to this long-standing problem in international nuclear safeguards.

Simulator test to study hot-flow problems related to a gas cooled reactor

NASA Technical Reports Server (NTRS)

Poole, J. W.; Freeman, M. P.; Doak, K. W.; Thorpe, M. L.

1973-01-01

An advance study of materials, fuel injection, and hot flow problems related to the gas core nuclear rocket is reported. The first task was to test a previously constructed induction heated plasma GCNR simulator above 300 kW. A number of tests are reported operating in the range of 300 kW at 10,000 cps. A second simulator was designed but not constructed for cold-hot visualization studies using louvered walls. A third task was a paper investigation of practical uranium feed systems, including a detailed discussion of related problems. The last assignment resulted in two designs for plasma nozzle test devices that could be operated at 200 atm on hydrogen.

User Guidelines and Best Practices for CASL VUQ Analysis Using Dakota.

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

Adams, Brian M.; Coleman, Kayla; Hooper, Russell

2016-11-01

Sandia's Dakota software (available at http://dakota.sandia.gov) supports science and engineering transformation through advanced exploration of simulations. Specifically, it manages and analyzes ensembles of simulations to provide broader and deeper perspective for analysts and decision makers. This enables them to enhance understanding of risk, improve products, and assess simulation credibility. This manual offers Consortium for Advanced Simulation of Light Water Reactors (LWRs) (CASL) partners a guide to conducting Dakota-based VUQ studies for CASL problems. It motivates various classes of Dakota methods and includes examples of their use on representative application problems. On reading, a CASL analyst should understand why and howmore » to apply Dakota to a simulation problem.« less

Power Quality Improvement by Unified Power Quality Conditioner Based on CSC Topology Using Synchronous Reference Frame Theory

PubMed Central

Dharmalingam, Rajasekaran; Dash, Subhransu Sekhar; Senthilnathan, Karthikrajan; Mayilvaganan, Arun Bhaskar; Chinnamuthu, Subramani

2014-01-01

This paper deals with the performance of unified power quality conditioner (UPQC) based on current source converter (CSC) topology. UPQC is used to mitigate the power quality problems like harmonics and sag. The shunt and series active filter performs the simultaneous elimination of current and voltage problems. The power fed is linked through common DC link and maintains constant real power exchange. The DC link is connected through the reactor. The real power supply is given by the photovoltaic system for the compensation of power quality problems. The reference current and voltage generation for shunt and series converter is based on phase locked loop and synchronous reference frame theory. The proposed UPQC-CSC design has superior performance for mitigating the power quality problems. PMID:25013854

Power quality improvement by unified power quality conditioner based on CSC topology using synchronous reference frame theory.

PubMed

Dharmalingam, Rajasekaran; Dash, Subhransu Sekhar; Senthilnathan, Karthikrajan; Mayilvaganan, Arun Bhaskar; Chinnamuthu, Subramani

2014-01-01

This paper deals with the performance of unified power quality conditioner (UPQC) based on current source converter (CSC) topology. UPQC is used to mitigate the power quality problems like harmonics and sag. The shunt and series active filter performs the simultaneous elimination of current and voltage problems. The power fed is linked through common DC link and maintains constant real power exchange. The DC link is connected through the reactor. The real power supply is given by the photovoltaic system for the compensation of power quality problems. The reference current and voltage generation for shunt and series converter is based on phase locked loop and synchronous reference frame theory. The proposed UPQC-CSC design has superior performance for mitigating the power quality problems.

In Situ NDA Conformation Measurements Performed at Auxiliary Charcoal Bed and Other Main Charcoal Beds After Uranium Removal from Molten Salt Reactor Experiment ACB at Oak Ridge National Laboratory

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

Haghighi, M. H.; Kring, C. T.; McGehee, J. T.

2002-02-26

The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR). The MSRE was run by Oak Ridge National Laboratory (ORNL) to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503. The reactor was operated from June 1965 tomore » December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed to cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. Beginning in 1987, it was discovered that gaseous uranium (U-233/U-232) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 had been generated when radiolysis in the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine. Some of the free fluorine combined with uranium fluorides (UF4) in the salt to produce UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE. One of the systems that UF6 migrated into due to this process was the offgas system which is vented to the MSRE main charcoal beds and MSRE auxiliary charcoal bed (ACB). Recently, the majority of the uranium laden-charcoal material residing within the ACB was safely and successfully removed using the uranium deposit removal system and equipment. After removal a series of NDA measurements was performed to determine the amount of uranium material remaining in the ACB, the amount of uranium material removed from the ACB, and the amount of uranium material remaining in the uranium removal equipment due to removal activities.« less

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

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

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in thismore » paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.« less

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

DOE PAGES

Betzler, Benjamin R.; Powers, Jeffrey J.; Worrall, Andrew

2017-03-01

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in thismore » paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.« less

Modified rotating biological contactor for removal of dichloromethane vapours.

PubMed

Ravi, R; Philip, Ligy; Swaminathan, T

2015-01-01

Bioreactors are used for the treatment of waste gas and odour that has gained much acceptance in the recent years to treat volatile organic compounds (VOCs). The different types of bioreactors (biofilter, biotrickling filter and bioscrubber) have been used for waste gas treatment. Each of these reactors has some advantages and some limitations. Though biodegradation is the main process for the removal of the pollutants, the mechanisms of removal and the microbial communities may differ among these bioreactors. Consequently, their performance or removal efficiency may also be different. Clogging of reactor and pressure drop are the main problems. In this study attempts are made to use the principle of rotating biological contactor (RBC) used for wastewater treatment for the removal of VOC. To overcome the above problem the RBC is modified which is suitable for the treatment of VOC (dichloromethane, DCM). DCM is harmful to human health and hazardous to the atmospheric environment. Modified RBC had no clogging problems and no pressure drop. So, it can handle the pollutant load for a longer period of time. A maximum elimination capacity of 25.7 g/m3 h has been achieved in this study for the DCM inlet load of 58 g/m3 h. The average biofilm thickness is 1 mm. The transient behaviour of the modified RBC treating DCM was investigated. The modified RBC is able to handle shutdown, restart and shock loading operations.

Results of comparative RBMK neutron computation using VNIIEF codes (cell computation, 3D statics, 3D kinetics). Final report

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

Grebennikov, A.N.; Zhitnik, A.K.; Zvenigorodskaya, O.A.

1995-12-31

In conformity with the protocol of the Workshop under Contract {open_quotes}Assessment of RBMK reactor safety using modern Western Codes{close_quotes} VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEUmore » codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core.« less

On theoretical and experimental modeling of metabolism forming in prebiotic systems

NASA Astrophysics Data System (ADS)

Bartsev, S. I.; Mezhevikin, V. V.

Recently searching for extraterrestrial life attracts more and more attention However the searching hardly can be effective without sufficiently universal concept of life origin which incidentally tackles a problem of origin of life on the Earth A concept of initial stages of life origin including origin of prebiotic metabolism is stated in the paper Suggested concept eliminates key difficulties in the problem of life origin and allows experimental verification of it According to the concept the predecessor of living beings has to be sufficiently simple to provide non-zero probability of self-assembling during short in geological or cosmic scale time In addition the predecessor has to be capable of autocatalysis and further complication evolution A possible scenario of initial stage of life origin which can be realized both on other planets and inside experimental facility is considered In the scope of the scenario a theoretical model of multivariate oligomeric autocatalyst coupled with phase-separated particle is presented Results of computer simulation of possible initial stage of chemical evolution are shown Conducted estimations show the origin of autocatalytic oligomeric phase-separated system is possible at reasonable values of kinetic parameters of involved chemical reactions in a small-scale flow reactor Accepted statements allowing to eliminate key problems of life origin imply important consequence -- organisms emerged out of the Earth or inside a reactor have to be based on another different from terrestrial biochemical

Attrition resistant catalysts for slurry-phase Fischer-Tropsch process

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

K. Jothimurugesan

1999-11-01

The Fischer-Tropsch (F-T) reaction provides a way of converting coal-derived synthesis gas (CO+H{sub 2}) to liquid fuels. Since the reaction is highly exothermic, one of the major problems in control of the reaction is heat removal. Recent work has shown that the use of slurry bubble column reactors (SBCRs) can largely solve this problem. Iron-based (Fe) catalysts are preferred catalysts for F-T because they are relatively inexpensive and possess reasonable activity for F-T synthesis (FTS). Their most advantages trait is their high water-gas shift (WGS) activity compared to their competitor, namely cobalt. This enables Fe F-T catalysts to process lowmore » H{sub 2}/CO ratio synthesis gas without an external shift reaction step. However, a serious problem with the use of Fe catalysts in a SBCR is their tendency to undergo attrition. This can cause fouling/plugging of downstream filters and equipment, make the separation of catalyst from the oil/wax product very difficult if not impossible, an d result in a steady loss of catalyst from the reactor. The objectives of this research were to develop a better understanding of the parameters affecting attrition of Fe F-T catalysts suitable for use in SBCRs and to incorporate this understanding into the design of novel Fe catalysts having superior attrition resistance.« less

RADIOACTIVE CONTAMINATION OF FOODS. PROBLEMS IN THE FOOD CONSUMPTION OF THE ITALIAN POPULATION

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

Ferro-Luzzi, A.; Mariani, A.

The aspects of health physics that are basically applications of physics are reviewed. Units of radiation measurement, RBE, permissible doses, personnel monitoring, applications of radiation spectrometry, and measurement of body activity are considered, as well as the release, dispersion, and deposition of radioactive material in reactor accidents. 140 references. (D.C.W.)

Oxygen-Induced Cracking Distillation of Oil in the Continuous Flow Tank Reactor

ERIC Educational Resources Information Center

Shvets, Valeriy F.; Kozlovskiy, Roman A.; Luganskiy, Artur I.; Gorbunov, Andrey V.; Suchkov, Yuriy P.; Ushin, Nikolay S.; Cherepanov, Alexandr A.

2016-01-01

The article analyses problems of processing black oil fuel and addresses the possibility of increasing the depth of oil refining by a new processing scheme. The study examines various methods of increasing the depth of oil refining reveals their inadequacies and highlights a need to introduce a new method of processing atmospheric and vacuum…

Multiphysics analysis of liquid metal annular linear induction pumps: A project overview

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

Maidana, Carlos Omar; Nieminen, Juha E.

Liquid metal-cooled fission reactors are both moderated and cooled by a liquid metal solution. These reactors are typically very compact and they can be used in regular electric power production, for naval and space propulsion systems or in fission surface power systems for planetary exploration. The coupling between the electromagnetics and thermo-fluid mechanical phenomena observed in liquid metal thermo-magnetic systems for nuclear and space applications gives rise to complex engineering magnetohydrodynamics and numerical problems. It is known that electromagnetic pumps have a number of advantages over rotating mechanisms: absence of moving parts, low noise and vibration level, simplicity of flowmore » rate regulation, easy maintenance and so on. However, while developing annular linear induction pumps, we are faced with a significant problem of magnetohydrodynamic instability arising in the device. The complex flow behavior in this type of devices includes a time-varying Lorentz force and pressure pulsation due to the time-varying electromagnetic fields and the induced convective currents that originates from the liquid metal flow, leading to instability problems along the device geometry. The determinations of the geometry and electrical configuration of liquid metal thermo-magnetic devices give rise to a complex inverse magnetohydrodynamic field problem were techniques for global optimization should be used, magnetohydrodynamics instabilities understood –or quantified- and multiphysics models developed and analyzed. Lastly, we present a project overview as well as a few computational models developed to study liquid metal annular linear induction pumps using first principles and the a few results of our multi-physics analysis.« less

Multiphysics analysis of liquid metal annular linear induction pumps: A project overview

DOE PAGES

Maidana, Carlos Omar; Nieminen, Juha E.

2016-03-14

Liquid metal-cooled fission reactors are both moderated and cooled by a liquid metal solution. These reactors are typically very compact and they can be used in regular electric power production, for naval and space propulsion systems or in fission surface power systems for planetary exploration. The coupling between the electromagnetics and thermo-fluid mechanical phenomena observed in liquid metal thermo-magnetic systems for nuclear and space applications gives rise to complex engineering magnetohydrodynamics and numerical problems. It is known that electromagnetic pumps have a number of advantages over rotating mechanisms: absence of moving parts, low noise and vibration level, simplicity of flowmore » rate regulation, easy maintenance and so on. However, while developing annular linear induction pumps, we are faced with a significant problem of magnetohydrodynamic instability arising in the device. The complex flow behavior in this type of devices includes a time-varying Lorentz force and pressure pulsation due to the time-varying electromagnetic fields and the induced convective currents that originates from the liquid metal flow, leading to instability problems along the device geometry. The determinations of the geometry and electrical configuration of liquid metal thermo-magnetic devices give rise to a complex inverse magnetohydrodynamic field problem were techniques for global optimization should be used, magnetohydrodynamics instabilities understood –or quantified- and multiphysics models developed and analyzed. Lastly, we present a project overview as well as a few computational models developed to study liquid metal annular linear induction pumps using first principles and the a few results of our multi-physics analysis.« less

The MSFR as a flexible CR reactor: the viewpoint of safety

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

Fiorina, C.; Cammi, A.; Franceschini, F.

2013-07-01

In this paper, the possibility has first been discussed of using the liquid-fuelled Molten Salt Fast Reactor (MSFR) as a flexible conversion ratio (CR) reactor without design modification. By tuning the reprocessing rate it is possible to determine the content of fission products in the core, which in turn can significantly affect the neutron economy without incurring in solubility problems. The MSFR can thus be operated as U-233 breeder (CR>1), iso-breeder (CR=1) and burner reactor (CR<1). In particular a 40 year doubling time can be achieved, as well as a considerable Transuranics and MA (minor actinide) burning rate equal tomore » about 150 kg{sub HN}/GWE-yr. The safety parameters of the MSFR have then been evaluated for different fuel cycle strategies. Th use and a softer spectrum combine to give a strong Doppler coefficient, one order of magnitude higher compared to traditional fast reactors (FRs). The fuel expansion coefficient is comparable to the Doppler coefficient and is only mildly affected by core compositions, thus assisting the fuel cycle flexibility of the MSFR. βeff and generation time are comparable to the case of traditional FRs, if a static fuel is assumed. A notable reduction of βeff is caused by salt circulation, but a low value of this parameter is a limited concern in the MSFR thanks to the lack of a burnup reactivity swing and of positive feedbacks. A simple approach has also been developed to evaluate the MSFR capabilities to withstand all typical double-fault accidents, for different fuel cycle options.« less

Potential Applications of Zeolite Membranes in Reaction Coupling Separation Processes

PubMed Central

Daramola, Michael O.; Aransiola, Elizabeth F.; Ojumu, Tunde V.

2012-01-01

Future production of chemicals (e.g., fine and specialty chemicals) in industry is faced with the challenge of limited material and energy resources. However, process intensification might play a significant role in alleviating this problem. A vision of process intensification through multifunctional reactors has stimulated research on membrane-based reactive separation processes, in which membrane separation and catalytic reaction occur simultaneously in one unit. These processes are rather attractive applications because they are potentially compact, less capital intensive, and have lower processing costs than traditional processes. Therefore this review discusses the progress and potential applications that have occurred in the field of zeolite membrane reactors during the last few years. The aim of this article is to update researchers in the field of process intensification and also provoke their thoughts on further research efforts to explore and exploit the potential applications of zeolite membrane reactors in industry. Further evaluation of this technology for industrial acceptability is essential in this regard. Therefore, studies such as techno-economical feasibility, optimization and scale-up are of the utmost importance.

Evaluation of different types of anaerobic seed sludge for the high rate anaerobic digestion of pig slurry in UASB reactors.

PubMed

Rico, Carlos; Montes, Jesús A; Rico, José Luis

2017-08-01

Three different types of anaerobic sludge (granular, thickened digestate and anaerobic sewage) were evaluated as seed inoculum sources for the high rate anaerobic digestion of pig slurry in UASB reactors. Granular sludge performance was optimal, allowing a high efficiency process yielding a volumetric methane production rate of 4.1LCH 4 L -1 d -1 at 1.5days HRT (0.248LCH 4 g -1 COD) at an organic loading rate of 16.4gCODL -1 d -1 . The thickened digestate sludge experimented flotation problems, thus resulting inappropriate for the UASB process. The anaerobic sewage sludge reactor experimented biomass wash-out, but allowed high process efficiency operation at 3days HRT, yielding a volumetric methane production rate of 1.7LCH 4 L -1 d -1 (0.236LCH 4 g -1 COD) at an organic loading rate of 7.2gCODL -1 d -1 . To guarantee the success of the UASB process, the settleable solids of the slurry must be previously removed. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

PubMed Central

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

2014-01-01

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

EDF experience with {open_quotes}hot spot{close_quotes} management

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

Guio, J.M. de

1995-03-01

During the past few years, {open_quotes}hot spots{close_quotes} due to the presence of particles of metal activated during their migration through the reactor core, have been detected at several French pressurized water reactor (PWR) units. These {open_quotes}hot spots,{close_quotes} which generate very high dose rates (from about 10 Gy/h to 200 G/h) are a significant factor in increase occupational exposures during outrates. Of particular concern are the difficult cases which prolong outage duration and increase the volume of radiological waste. Confronted with this situation, Electricite de France (EDF) has set up a national research group, as part of its ALARA program, tomore » establish procedures and techniques to avoid, detect, and eliminate of hot spots. In particular, specific processes have been developed to eliminate these hot spots which are most costly in terms of occupational exposure due to the need for reactor maintenance. This paper sets out the general approach adopted at EDF so far to cope with the problem of hot spots, illustrated by experience at Blayais 3 and 4.« less

Application of gaseous core reactors for transmutation of nuclear waste

NASA Technical Reports Server (NTRS)

Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

1976-01-01

An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

Minimization of the energy loss of nuclear power plants in case of partial in-core monitoring system failure

NASA Astrophysics Data System (ADS)

Zagrebaev, A. M.; Ramazanov, R. N.; Lunegova, E. A.

2017-01-01

In this paper we consider the optimization problem minimize of the energy loss of nuclear power plants in case of partial in-core monitoring system failure. It is possible to continuation of reactor operation at reduced power or total replacement of the channel neutron measurements, requiring shutdown of the reactor and the stock of detectors. This article examines the reconstruction of the energy release in the core of a nuclear reactor on the basis of the indications of height sensors. The missing measurement information can be reconstructed by mathematical methods, and replacement of the failed sensors can be avoided. It is suggested that a set of ‘natural’ functions determined by means of statistical estimates obtained from archival data be constructed. The procedure proposed makes it possible to reconstruct the field even with a significant loss of measurement information. Improving the accuracy of the restoration of the neutron flux density in partial loss of measurement information to minimize the stock of necessary components and the associated losses.

Burn Control in Fusion Reactors via Isotopic Fuel Tailoring

NASA Astrophysics Data System (ADS)

Boyer, Mark D.; Schuster, Eugenio

2011-10-01

The control of plasma density and temperature are among the most fundamental problems in fusion reactors and will be critical to the success of burning plasma experiments like ITER. Economic and technological constraints may require future commercial reactors to operate with low temperature, high-density plasma, for which the burn condition may be unstable. An active control system will be essential for stabilizing such operating points. In this work, a volume-averaged transport model for the energy and the densities of deuterium and tritium fuel ions, as well as the alpha particles, is used to synthesize a nonlinear feedback controller for stabilizing the burn condition. The controller makes use of ITER's planned isotopic fueling capability and controls the densities of these ions separately. The ability to modulate the DT fuel mix is exploited in order to reduce the fusion power during thermal excursions without the need for impurity injection. By moving the isotopic mix in the plasma away from the optimal 50:50 mix, the reaction rate is slowed and the alpha-particle heating is reduced to desired levels. Supported by the NSF CAREER award program (ECCS-0645086).

History of fast reactor fuel development

NASA Astrophysics Data System (ADS)

Kittel, J. H.; Frost, B. R. T.; Mustelier, J. P.; Bagley, K. Q.; Crittenden, G. C.; Van Dievoet, J.

1993-09-01

The first fast breeder reactors, constructed in the 1945-1960 time period, used metallic fuels composed of uranium, plutonium, or their alloys. They were chosen because most existing reactor operating experience had been obtained on metallic fuels and because they provided the highest breeding ratios. Difficulties in obtaining adequate dimensional stability in metallic fuel elements under conditions of high fuel burnup led in the 1960s to the virtual worldwide choice of ceramic fuels. Although ceramic fuels provide lower breeding performance, this objective is no longer an important consideration in most national programs. Mixed uranium and plutonium dioxide became the ceramic fuel that has received the widest use. The more advanced ceramic fuels, mixed uranium and plutonium carbides and nitrides, continue under development. More recently, metal fuel elements of improved design have joined ceramic fuels in achieving goal burnups of 15 to 20 percent. Low-swelling fuel cladding alloys have also been continuously developed to deal with the unexpected problem of void formation in stainless steels subjected to fast neutron irradiation, a phenomenon first observed in the 1960s.

Effects of turbulence modelling on prediction of flow characteristics in a bench-scale anaerobic gas-lift digester.

PubMed

Coughtrie, A R; Borman, D J; Sleigh, P A

2013-06-01

Flow in a gas-lift digester with a central draft-tube was investigated using computational fluid dynamics (CFD) and different turbulence closure models. The k-ω Shear-Stress-Transport (SST), Renormalization-Group (RNG) k-∊, Linear Reynolds-Stress-Model (RSM) and Transition-SST models were tested for a gas-lift loop reactor under Newtonian flow conditions validated against published experimental work. The results identify that flow predictions within the reactor (where flow is transitional) are particularly sensitive to the turbulence model implemented; the Transition-SST model was found to be the most robust for capturing mixing behaviour and predicting separation reliably. Therefore, Transition-SST is recommended over k-∊ models for use in comparable mixing problems. A comparison of results obtained using multiphase Euler-Lagrange and singlephase approaches are presented. The results support the validity of the singlephase modelling assumptions in obtaining reliable predictions of the reactor flow. Solver independence of results was verified by comparing two independent finite-volume solvers (Fluent-13.0sp2 and OpenFOAM-2.0.1). Copyright © 2013 Elsevier Ltd. All rights reserved.

ROMUSE 2.0 User Manual

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

Khuwaileh, Bassam; Turinsky, Paul; Williams, Brian J.

2016-10-04

ROMUSE (Reduced Order Modeling Based Uncertainty/Sensitivity Estimator) is an effort within the Consortium for Advanced Simulation of Light water reactors (CASL) to provide an analysis tool to be used in conjunction with reactor core simulators, especially the Virtual Environment for Reactor Applications (VERA). ROMUSE is written in C++ and is currently capable of performing various types of parameters perturbations, uncertainty quantification, surrogate models construction and subspace analysis. Version 2.0 has the capability to interface with DAKOTA which gives ROMUSE access to the various algorithms implemented within DAKOTA. ROMUSE is mainly designed to interface with VERA and the Comprehensive Modeling andmore » Simulation Suite for Nuclear Safety Analysis and Design (SCALE) [1,2,3], however, ROMUSE can interface with any general model (e.g. python and matlab) with Input/Output (I/O) format that follows the Hierarchical Data Format 5 (HDF5). In this brief user manual, the use of ROMUSE will be overviewed and example problems will be presented and briefly discussed. The algorithms provided here range from algorithms inspired by those discussed in Ref.[4] to nuclear-specific algorithms discussed in Ref. [3].« less

Scaling of surface-plasma reactors with a significantly increased energy density for NO conversion.

PubMed

Malik, Muhammad Arif; Xiao, Shu; Schoenbach, Karl H

2012-03-30

Comparative studies revealed that surface plasmas developing along a solid-gas interface are significantly more effective and energy efficient for remediation of toxic pollutants in air than conventional plasmas propagating in air. Scaling of the surface plasma reactors to large volumes by operating them in parallel suffers from a serious problem of adverse effects of the space charges generated at the dielectric surfaces of the neighboring discharge chambers. This study revealed that a conductive foil on the cathode potential placed between the dielectric plates as a shield not only decoupled the discharges, but also increased the electrical power deposited in the reactor by a factor of about forty over the electrical power level obtained without shielding and without loss of efficiency for NO removal. The shield had no negative effect on efficiency, which is verified by the fact that the energy costs for 50% NO removal were about 60 eV/molecule and the energy constant, k(E), was about 0.02 L/J in both the shielded and unshielded cases. Copyright © 2012 Elsevier B.V. All rights reserved.

FY2012 summary of tasks completed on PROTEUS-thermal work.

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

Lee, C.H.; Smith, M.A.

2012-06-06

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targetedmore » reactor types. The DeCART code has been reorganized in preparation for an efficient integration of the different versions for PWR, BWR, and VHTR analysis. In DeCART, the old-fashioned common blocks and header files have been replaced by advanced memory structures. However, the changing of variable names was minimized in order to limit problems with the code integration. Since the remaining stability problems of DeCART were mostly caused by the CMFD methodology and modules, significant work was performed to determine whether they could be replaced by more stable methods and routines. The cross section library is a key element to obtain accurate solutions. Thus, the procedure for generating cross section libraries was revisited to provide libraries tailored for the targeted reactor types. To improve accuracy in the cross section library, an attempt was made to replace the CENTRM code by the MCNP Monte Carlo code as a tool obtaining reference resonance integrals. The use of the Monte Carlo code allows us to minimize problems or approximations that CENTRM introduces since the accuracy of the subgroup data is limited by that of the reference solutions. The use of MCNP requires an additional set of libraries without resonance cross sections so that reference calculations can be performed for a unit cell in which only one isotope of interest includes resonance cross sections, among the isotopes in the composition. The OECD MHTGR-350 benchmark core was simulated using DeCART as initial focus of the verification/validation efforts. Among the benchmark problems, Exercise 1 of Phase 1 is a steady-state benchmark case for the neutronics calculation for which block-wise cross sections were provided in 26 energy groups. This type of problem was designed for a homogenized geometry solver like DIF3D rather than the high-fidelity code DeCART. Instead of the homogenized block cross sections given in the benchmark, the VHTR-specific 238-group ENDF/B-VII.0 library of DeCART was directly used for preliminary calculations. Initial results showed that the multiplication factors of a fuel pin and a fuel block with or without a control rod hole were off by 6, -362, and -183 pcm Dk from comparable MCNP solutions, respectively. The 2-D and 3-D one-third core calculations were also conducted for the all-rods-out (ARO) and all-rods-in (ARI) configurations, producing reasonable results. Figure 1 illustrates the intermediate (1.5 eV - 17 keV) and thermal (below 1.5 eV) group flux distributions. As seen from VHTR cores with annular fuels, the intermediate group fluxes are relatively high in the fuel region, but the thermal group fluxes are higher in the inner and outer graphite reflector regions than in the fuel region. To support the current project, a new three-year I-NERI collaboration involving ANL and KAERI was started in November 2011, focused on performing in-depth verification and validation of high-fidelity multi-physics simulation codes for LWR and VHTR. The work scope includes generating improved cross section libraries for the targeted reactor types, developing benchmark models for verification and validation of the neutronics code with or without thermo-fluid feedback, and performing detailed comparisons of predicted reactor parameters against both Monte Carlo solutions and experimental measurements. The following list summarizes the work conducted so far for PROTEUS-Thermal Tasks: Unification of different versions of DeCART was initiated, and at the same time code modernization was conducted to make code unification efficient; (2) Regeneration of cross section libraries was attempted for the targeted reactor types, and the procedure for generating cross section libraries was updated by replacing CENTRM with MCNP for reference resonance integrals; (3) The MHTGR-350 benchmark core was simulated using DeCART with VHTR-specific 238-group ENDF/B-VII.0 library, and MCNP calculations were performed for comparison; and (4) Benchmark problems for PWR and BWR analysis were prepared for the DeCART verification/validation effort. In the coming months, the work listed above will be completed. Cross section libraries will be generated with optimized group structures for specific reactor types.« less

Implementation of the SPH Procedure Within the MOOSE Finite Element Framework

NASA Astrophysics Data System (ADS)

Laurier, Alexandre

The goal of this thesis was to implement the SPH homogenization procedure within the MOOSE finite element framework at INL. Before this project, INL relied on DRAGON to do their SPH homogenization which was not flexible enough for their needs. As such, the SPH procedure was implemented for the neutron diffusion equation with the traditional, Selengut and true Selengut normalizations. Another aspect of this research was to derive the SPH corrected neutron transport equations and implement them in the same framework. Following in the footsteps of other articles, this feature was implemented and tested successfully with both the PN and S N transport calculation schemes. Although the results obtained for the power distribution in PWR assemblies show no advantages over the use of the SPH diffusion equation, we believe the inclusion of this transport correction will allow for better results in cases where either P N or SN are required. An additional aspect of this research was the implementation of a novel way of solving the non-linear SPH problem. Traditionally, this was done through a Picard, fixed-point iterative process whereas the new implementation relies on MOOSE's Preconditioned Jacobian-Free Newton Krylov (PJFNK) method to allow for a direct solution to the non-linear problem. This novel implementation showed a decrease in calculation time by a factor reaching 50 and generated SPH factors that correspond to those obtained through a fixed-point iterative process with a very tight convergence criteria: epsilon < 10-8. The use of the PJFNK SPH procedure also allows to reach convergence in problems containing important reflector regions and void boundary conditions, something that the traditional SPH method has never been able to achieve. At times when the PJFNK method cannot reach convergence to the SPH problem, a hybrid method is used where by the traditional SPH iteration forces the initial condition to be within the radius of convergence of the Newton method. This new method was tested on a simplified model of INL's TREAT reactor, a problem that includes very important graphite reflector regions as well as vacuum boundary conditions with great success. To demonstrate the power of PJFNK SPH on a more common case, the correction was applied to a simplified PWR reactor core from the BEAVRS benchmark that included 15 assemblies and the water reflector to obtain very good results. This opens up the possibility to apply the SPH correction to full reactor cores in order to reduce homogenization errors for use in transient or multi-physics calculations.

High-efficiency removal of NOx using dielectric barrier discharge nonthermal plasma with water as an outer electrode

NASA Astrophysics Data System (ADS)

Dan, ZHAO; Feng, YU; Amin, ZHOU; Cunhua, MA; Bin, DAI

2018-01-01

With the rapid increase in the number of cars and the development of industry, nitrogen oxide (NOx) emissions have become a serious and pressing problem. This work reports on the development of a water-cooled dielectric barrier discharge reactor for gaseous NOx removal at low temperature. The characteristics of the reactor are evaluated with and without packing of the reaction tube with 2 mm diameter dielectric beads composed of glass, ZnO, MnO2, ZrO2, or Fe2O3. It is found that the use of a water-cooled tube reduces the temperature, which stabilizes the reaction, and provides a much greater NO conversion efficiency (28.8%) than that obtained using quartz tube (14.1%) at a frequency of 8 kHz with an input voltage of 6.8 kV. Furthermore, under equivalent conditions, packing the reactor tube with glass beads greatly increases the NO conversion efficiency to 95.85%. This is because the dielectric beads alter the distribution of the electric field due to the influence of polarization at the glass bead surfaces, which ultimately enhances the plasma discharge intensity. The presence of the dielectric beads increases the gas residence time within the reactor. Experimental verification and a theoretical basis are provided for the industrial application of the proposed plasma NO removal process employing dielectric bead packing.

Using Coupled Mesoscale Experiments and Simulations to Investigate High Burn-Up Oxide Fuel Thermal Conductivity

NASA Astrophysics Data System (ADS)

Teague, Melissa C.; Fromm, Bradley S.; Tonks, Michael R.; Field, David P.

2014-12-01

Nuclear energy is a mature technology with a small carbon footprint. However, work is needed to make current reactor technology more accident tolerant and to allow reactor fuel to be burned in a reactor for longer periods of time. Optimizing the reactor fuel performance is essentially a materials science problem. The current understanding of fuel microstructure have been limited by the difficulty in studying the structure and chemistry of irradiated fuel samples at the mesoscale. Here, we take advantage of recent advances in experimental capabilities to characterize the microstructure in 3D of irradiated mixed oxide (MOX) fuel taken from two radial positions in the fuel pellet. We also reconstruct these microstructures using Idaho National Laboratory's MARMOT code and calculate the impact of microstructure heterogeneities on the effective thermal conductivity using mesoscale heat conduction simulations. The thermal conductivities of both samples are higher than the bulk MOX thermal conductivity because of the formation of metallic precipitates and because we do not currently consider phonon scattering due to defects smaller than the experimental resolution. We also used the results to investigate the accuracy of simple thermal conductivity approximations and equations to convert 2D thermal conductivities to 3D. It was found that these approximations struggle to predict the complex thermal transport interactions between metal precipitates and voids.

Ab initio Investigation of Helium in Vanadium Oxide Nanoclusters

NASA Astrophysics Data System (ADS)

Danielson, Thomas; Tea, Eric; Hin, Celine

Nanostructured ferritic alloys (NFAs) are strong candidate materials for the next generation of fission reactors and future fusion reactors. They are characterized by a large number density of oxide nanoclusters dispersed throughout a BCC iron matrix, where current oxide nanoclusters are primarily comprised of Y-Ti-O compounds. The oxide nanoclusters provide the alloy with high resistance to neutron irradiation, high yield strength and high creep strength at the elevated temperatures of a reactor environment. In addition, the oxide nanoclusters serve as trapping sites for transmutation product helium providing substantially increased resistance to catastrophic cracking and embrittlement. Although the mechanical properties and radiation resistance of the existing NFAs is promising, the problem of forming large scale reactor components continues to present a formidable challenge due to the high hardness and unpredictable fracture behavior of the alloys. An alternative alloy has been previously proposed and fabricated where vanadium is added in order to form vanadium oxide nanoclusters that serve as deflection sites for crack propagation. Although experiments have shown evidence that the fracture behavior of the alloys is improved, it is unknown whether or not the vanadium oxide nanoclusters are effective trapping sites for helium. We present results obtained using density functional theory investigating the thermodynamic stability of helium with the vanadium oxide matrix to make a comparison of trapping effectiveness to traditional Y-Ti-O compounds.

MODFLOW 2.0: A program for predicting moderator flow patterns

NASA Astrophysics Data System (ADS)

Peterson, P. F.; Paik, I. K.

1991-07-01

Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in the operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides transient moderator flow pattern information with stratification effects, and tracks the location of ink plumes in the reactor. The code, written in Fortran, is compiled for Macintosh II computers, and includes subroutines for interactive control and graphical output. Removing the graphics capabilities, the code can also be compiled on other computers. With graphics, in addition to the capability to perform safety related computations, MODFLOW also provides an easy tool for becoming familiar with flow distributions in SRS reactors.

Reactivity Insertion Accident (RIA) Capability Status in the BISON Fuel Performance Code

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

Williamson, Richard L.; Folsom, Charles Pearson; Pastore, Giovanni

2016-05-01

One of the Challenge Problems being considered within CASL relates to modelling and simulation of Light Water Reactor LWR) fuel under Reactivity Insertion Accident (RIA) conditions. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions, and thus must be capable of addressing the RIA challenge problem. This report outlines required BISON capabilities for RIAs and describes the current status of the code. Information on recent accident capability enhancements, application of BISON to a RIA benchmark exercise, and plans for validation to RIA behavior are included.

Telling the public about risks

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

Ahearne, J.F.

1990-09-01

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

On-line infrared process signature measurements through combustion atmospheres

NASA Astrophysics Data System (ADS)

Zweibaum, F. M.; Kozlowski, A. T.; Surette, W. E., Jr.

1980-01-01

A number of on-line infrared process signature measurements have been made through combustion atmospheres, including those in jet engines, piston engines, and coal gasification reactors. The difficulties involved include operation in the presence of pressure as high as 1800 psi, temperatures as high as 3200 F, and explosive, corrosive and dust-laden atmospheres. Calibration problems have resulted from the use of purge gases to clear the viewing tubes, and the obscuration of the view ports by combustion products. A review of the solutions employed to counteract the problems is presented, and areas in which better solutions are required are suggested.

Storage and treatment of SNF of Alfa class nuclear submarines: current status and problems

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

Ignatiev, Sviatoslav; Zabudko, Alexey; Pankratov, Dmitry

Available in abstract form only. Full text of publication follows: The current status and main problems associated with storage, defueling and following treatment of spent nuclear fuel (SNF) of Nuclear Submarines (NS) with heavy liquid metal cooled reactors are considered. In the final analysis these solutions could be realized in the form of separate projects to be funded through national and bi- and multilateral funding in the framework of the international collaboration of the Russian Federation on complex utilization of NS and rehabilitation of contaminated objects allocated in the North-West region of Russia. (authors)

Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

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

Peterson, Per; Greenspan, Ehud

2015-02-09

This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designsmore » are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel digital x-ray tomography methods to track both the translational and rotational motion of spherical pebbles, which provides unique experimental results that can be used to validate discrete element method (DEM) simulations of pebble motion. The validation effort supported by the X-PREX facility provides a means to build confidence in analysis of pebble bed configuration and residence time distributions that impact the neutronics, thermal hydraulics, and safety analysis of pebble bed reactor cores. Experimental and DEM simulation results are reported for silo drainage, a classical problem in the granular flow literature, at several hopper angles. These studies include conventional converging and novel diverging geometries that provide additional flexibility in the design of pebble bed reactor cores. Excellent agreement is found between the X-PREX experimental and DEM simulation results. This report also includes results for additional studies relevant to the design and analysis of pebble bed reactor cores including the study of forces on shut down blades inserted directly into a packed bed and pebble flow in a cylindrical hopper that is representative of a small test reactor.« less

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

DOE PAGES

Durham, J. M.; Poulson, D.; Bacon, J.; ...

2018-04-10

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

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

None

During this time period, at WVU, we tried several methods to eliminate problems related to condensation of heavier products when reduced Mo-Ni-K/C materials were used as catalysts. We then resumed our kinetic study on the reduced Mo-Ni-K/C catalysts. We have also obtained same preliminary results in our attempts to analyze quantitatively the temperature-programmed reduction (TPR) spectra for C-supported Mo-based catalysts. We have completed the kinetic study for the sulfided Co-K-MoS /C catalyst. We have compared the results of methanol synthesis 2 using the membrane reactor with those using a simple plug-flow reactor. At UCC, the complete characterization of selected catalystsmore » has been completed. The results suggest that catalyst pretreatment under different reducing conditions yield different surface compositions and thus different catalytic reactivities.« less

An Experimental Investigation of Sewage Sludge Gasification in a Fluidized Bed Reactor

PubMed Central

Calvo, L. F.; García, A. I.; Otero, M.

2013-01-01

The gasification of sewage sludge was carried out in a simple atmospheric fluidized bed gasifier. Flow and fuel feed rate were adjusted for experimentally obtaining an air mass : fuel mass ratio (A/F) of 0.2 < A/F < 0.4. Fuel characterization, mass and power balances, produced gas composition, gas phase alkali and ammonia, tar concentration, agglomeration tendencies, and gas efficiencies were assessed. Although accumulation of material inside the reactor was a main problem, this was avoided by removing and adding bed media along gasification. This allowed improving the process heat transfer and, therefore, gasification efficiency. The heating value of the produced gas was 8.4 MJ/Nm, attaining a hot gas efficiency of 70% and a cold gas efficiency of 57%. PMID:24453863

Verification of Spent Nuclear Fuel in Sealed Dry Storage Casks via Measurements of Cosmic-Ray Muon Scattering

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

Durham, J. M.; Poulson, D.; Bacon, J.

Most of the plutonium in the world resides inside spent nuclear reactor fuel rods. This high-level radioactive waste is commonly held in long-term storage within large, heavily shielded casks. Currently, international nuclear safeguards inspectors have no stand-alone method of verifying the amount of reactor fuel stored within a sealed cask. In this paper, we demonstrate experimentally that measurements of the scattering angles of cosmic-ray muons, which pass through a storage cask, can be used to determine if spent fuel assemblies are missing without opening the cask. Finally, this application of technology and methods commonly used in high-energy particle physics providesmore » a potential solution to this long-standing problem in international nuclear safeguards.« less

Future Scenarios for Fission Based Reactors

NASA Astrophysics Data System (ADS)

David, S.

2005-04-01

The coming century will see the exhaustion of standard fossil fuels, coal, gas and oil, which today represent 75% of the world energy production. Moreover, their use will have caused large-scale emission of greenhouse gases (GEG), and induced global climate change. This problem is exacerbated by a growing world energy demand. In this context, nuclear power is the only GEG-free energy source available today capable of responding significantly to this demand. Some scenarios consider a nuclear energy production of around 5 Gtoe in 2050, wich would represent a 20% share of the world energy supply. Present reactors generate energy from the fission of U-235 and require around 200 tons of natural Uranium to produce 1GWe.y of energy, equivalent to the fission of one ton of fissile material. In a scenario of a significant increase in nuclear energy generation, these standard reactors will consume the whole of the world's estimated Uranium reserves in a few decades. However, natural Uranium or Thorium ore, wich are not themselves fissile, can produce a fissile material after a neutron capture ( 239Pu and 233U respectively). In a breeder reactor, the mass of fissile material remains constant, and the fertile ore is the only material to be consumed. In this case, only 1 ton of natural ore is needed to produce 1GWe.y. Thus, the breeding concept allows optimal use of fertile ore and development of sustainable nuclear energy production for several thousand years into the future. Different sustainable nuclear reactor concepts are studied in the international forum "generation IV". Different types of coolant (Na, Pb and He) are studied for fast breeder reactors based on the Uranium cycle. The thermal Thorium cycle requires the use of a liquid fuel, which can be reprocessed online in order to extract the neutron poisons. This paper presents these different sustainable reactors, based on the Uranium or Thorium fuel cycles and will compare the different options in term of fissile inventory, capacity to be deployed, induced radiotoxicities, and R&D efforts.

Leasing of Nuclear Power Plants With Using Floating Technologies

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

Kuznetsov, Yu.N.; Gabaraev, B.A.; Reshetov, V.A.

2002-07-01

The proposal to organize and realize the international program on leasing of Nuclear Power Plant (NPP) reactor compartments is brought to the notice of potential partners. The proposal is oriented to the construction of new NPPs or to replacement of worked-out reactor units of the NPPs in operation on the sites situated near water area and to the use of afloat technologies for construction, mounting and transportation of reactor units as a Reactor Compartment Block Module (RCBM). According to the offered project the RCBM is fabricated in factory conditions at the largest Russian defense shipbuilding plant - State Unitary Enterprisemore » 'Industrial Association SEVMASHPREDPRIYATIE' (SEVMASH) in the city of Severodvinsk of the Arkhangelsk region. After completion of assembling, testing and preliminary licensing the RCBM is given buoyancy by means of hermetic sealing and using pontoons and barges. The RCBM delivery to the NPP site situated near water area is performed by sea route. The RCBM is brought to the place of its installation with the use of appropriate hydraulic structures (canals, shipping locks), then is lowered on the basement constructed beforehand and incorporated into NPP scheme, of which the components are installed in advance. Floating means can be detached from the RCBM and used repeatedly for other RCBMs. Further procedure of NPP commissioning and its operation is carried out according to traditional method by power company in the framework of RCBM leasing with enlisting the services of firm-manufacturer's specialists either to provide reactor plant operation and concomitant processes or to perform author's supervision of operation. After completion of lifetime and reactor unloading the RCBM is dismantled with using the same afloat technology and taken away from NPP site to sea area entirely, together with its structures (reactor vessel, heat exchangers, pumps, pipelines and other equipment). Then RCBM is transported by shipping route to a firm-manufacturer, for subsequent reprocessing, utilization and storage. Nuclear fuel and radioactive wastes are removed from NPP site also. Use of leasing method removes legal problems connected with the transportation of radioactive materials through state borders as the RCBM remains a property of the state-producer at all stages of its life cycle. (authors)« less

Stochastic modelling of turbulent combustion for design optimization of gas turbine combustors

NASA Astrophysics Data System (ADS)

Mehanna Ismail, Mohammed Ali

The present work covers the development and the implementation of an efficient algorithm for the design optimization of gas turbine combustors. The purpose is to explore the possibilities and indicate constructive suggestions for optimization techniques as alternative methods for designing gas turbine combustors. The algorithm is general to the extent that no constraints are imposed on the combustion phenomena or on the combustor configuration. The optimization problem is broken down into two elementary problems: the first is the optimum search algorithm, and the second is the turbulent combustion model used to determine the combustor performance parameters. These performance parameters constitute the objective and physical constraints in the optimization problem formulation. The examination of both turbulent combustion phenomena and the gas turbine design process suggests that the turbulent combustion model represents a crucial part of the optimization algorithm. The basic requirements needed for a turbulent combustion model to be successfully used in a practical optimization algorithm are discussed. In principle, the combustion model should comply with the conflicting requirements of high fidelity, robustness and computational efficiency. To that end, the problem of turbulent combustion is discussed and the current state of the art of turbulent combustion modelling is reviewed. According to this review, turbulent combustion models based on the composition PDF transport equation are found to be good candidates for application in the present context. However, these models are computationally expensive. To overcome this difficulty, two different models based on the composition PDF transport equation were developed: an improved Lagrangian Monte Carlo composition PDF algorithm and the generalized stochastic reactor model. Improvements in the Lagrangian Monte Carlo composition PDF model performance and its computational efficiency were achieved through the implementation of time splitting, variable stochastic fluid particle mass control, and a second order time accurate (predictor-corrector) scheme used for solving the stochastic differential equations governing the particles evolution. The model compared well against experimental data found in the literature for two different configurations: bluff body and swirl stabilized combustors. The generalized stochastic reactor is a newly developed model. This model relies on the generalization of the concept of the classical stochastic reactor theory in the sense that it accounts for both finite micro- and macro-mixing processes. (Abstract shortened by UMI.)

DMM: A MULTIGROUP, MULTIREGION ONE-SPACE-DIMENSIONAL COMPUTER PROGRAM USING NEUTRON DIFFUSION THEORY. PART II. DMM PROGRAM DESCRIPTION

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

Kavanagh, D.L.; Antchagno, M.J.; Egawa, E.K.

1960-12-31

Operating instructions are presented for DMM, a Remington Rand 1103A program using one-space-dimensional multigroup diffusion theory to calculate the reactivity or critical conditions and flux distribution of a multiregion reactor. Complete descriptions of the routines and problem input and output specifications are also included. (D.L.C.)

The Behavior of Matter under Nonequilibrium Conditions: Fundamental Aspects and Applications in Energy-oriented Problems: Progress Report for Period September 1984--November 1987

DOE R&D Accomplishments Database

Prigogine, I.

1987-10-07

This report briefly discusses progress on the following topics: state selection dynamics; polymerization under nonequilibrium conditions; inhomogeneous fluctuations in hydrodynamics and in completely mixed reactors; homoclinic bifurcations and mixed-mode oscillations; intrinsic randomness and spontaneous symmetry breaking in explosive systems; and microscopic means of irreversibility.

The Paucity Problem: Where Have All the Space Reactor Experiments Gone?

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

Bess, John D.; Marshall, Margaret A.

2016-10-01

The Handbooks of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) together contain a plethora of documented and evaluated experiments essential in the validation of nuclear data, neutronics codes, and modeling of various nuclear systems. Unfortunately, only a minute selection of handbook data (twelve evaluations) are of actual experimental facilities and mockups designed specifically for space nuclear research. There is a paucity problem, such that the multitude of space nuclear experimental activities performed in the past several decades have yet to be recovered and made available in such detail that themore » international community could benefit from these valuable historical research efforts. Those experiments represent extensive investments in infrastructure, expertise, and cost, as well as constitute significantly valuable resources of data supporting past, present, and future research activities. The ICSBEP and IRPhEP were established to identify and verify comprehensive sets of benchmark data; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data. See full abstract in attached document.« less

Flow instability in particle-bed nuclear reactors

NASA Astrophysics Data System (ADS)

Kerrebrock, Jack L.

The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded that operation at design exit temperature but at reduced power could be hazardous for such a reactor. But the calculations also show that an appropriate cold frit could very likely cure the instability. More definite conclusions must await calculations for specific designs.

Neutron dose rate analysis on HTGR-10 reactor using Monte Carlo code

NASA Astrophysics Data System (ADS)

Suwoto; Adrial, H.; Hamzah, A.; Zuhair; Bakhri, S.; Sunaryo, G. R.

2018-02-01

The HTGR-10 reactor is cylinder-shaped core fuelled with kernel TRISO coated fuel particles in the spherical pebble with helium cooling system. The outlet helium gas coolant temperature outputted from the reactor core is designed to 700 °C. One advantage HTGR type reactor is capable of co-generation, as an addition to generating electricity, the reactor was designed to produce heat at high temperature can be used for other processes. The spherical fuel pebble contains 8335 TRISO UO2 kernel coated particles with enrichment of 10% and 17% are dispersed in a graphite matrix. The main purpose of this study was to analysis the distribution of neutron dose rates generated from HTGR-10 reactors. The calculation and analysis result of neutron dose rate in the HTGR-10 reactor core was performed using Monte Carlo MCNP5v1.6 code. The problems of double heterogeneity in kernel fuel coated particles TRISO and spherical fuel pebble in the HTGR-10 core are modelled well with MCNP5v1.6 code. The neutron flux to dose conversion factors taken from the International Commission on Radiological Protection (ICRP-74) was used to determine the dose rate that passes through the active core, reflectors, core barrel, reactor pressure vessel (RPV) and a biological shield. The calculated results of neutron dose rate with MCNP5v1.6 code using a conversion factor of ICRP-74 (2009) for radiation workers in the radial direction on the outside of the RPV (radial position = 220 cm from the center of the patio HTGR-10) provides the respective value of 9.22E-4 μSv/h and 9.58E-4 μSv/h for enrichment 10% and 17%, respectively. The calculated values of neutron dose rates are compliant with BAPETEN Chairman’s Regulation Number 4 Year 2013 on Radiation Protection and Safety in Nuclear Energy Utilization which sets the limit value for the average effective dose for radiation workers 20 mSv/year or 10μSv/h. Thus the protection and safety for radiation workers to be safe from the radiation source has been fulfilled. From the result analysis, it can be concluded that the model of calculation result of neutron dose rate for HTGR-10 core has met the required radiation safety standards.

Flow instability in particle-bed nuclear reactors

NASA Technical Reports Server (NTRS)

Kerrebrock, Jack L.

1993-01-01

The particle-bed core offers mitigation of some of the problems of solid-core nuclear rocket reactors. Dividing the fuel elements into small spherical particles contained in a cylindrical bed through which the propellant flows radially, may reduce the thermal stress in the fuel elements, allowing higher propellant temperatures to be reached. The high temperature regions of the reactor are confined to the interior of cylindrical fuel assemblies, so most of the reactor can be relatively cool. This enables the use of structural and moderating materials which reduce the minimum critical size and mass of the reactor. One of the unresolved questions about this concept is whether the flow through the particle-bed will be well behaved, or will be subject to destructive flow instabilities. Most of the recent analyses of the stability of the particle-bed reactor have been extensions of the approach of Bussard and Delauer, where the bed is essentially treated as an array of parallel passages, so that the mass flow is continuous from inlet to outlet through any one passage. A more general three dimensional model of the bed is adopted, in which the fluid has mobility in three dimensions. Comparison of results of the earlier approach to the present one shows that the former does not accurately represent the stability at low Re. The more complete model presented should be capable of meeting this deficiency while accurately representing the effects of the cold and hot frits, and of heat conduction and radiation in the particle-bed. It can be extended to apply to the cylindrical geometry of particle-bed reactors without difficulty. From the exemplary calculations which were carried out, it can be concluded that a particle-bed without a cold frit would be subject to instability if operated at the high temperatures desired for nuclear rockets, and at power densities below about 4 megawatts per liter. Since the desired power density is about 40 megawatts per liter, it can be concluded that operation at design exit temperature but at reduced power could be hazardous for such a reactor. But the calculations also show that an appropriate cold frit could very likely cure the instability. More definite conclusions must await calculations for specific designs.

Pyrolysis of waste tyres: A review

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

Williams, Paul T., E-mail: p.t.williams@leeds.ac.uk

2013-08-15

Graphical abstract: - Highlights: • Pyrolysis of waste tyres produces oil, gas and char, and recovered steel. • Batch, screw kiln, rotary kiln, vacuum and fluidised-bed are main reactor types. • Product yields are influenced by reactor type, temperature and heating rate. • Pyrolysis oils are complex and can be used as chemical feedstock or fuel. • Research into higher value products from the tyre pyrolysis process is reviewed. - Abstract: Approximately 1.5 billion tyres are produced each year which will eventually enter the waste stream representing a major potential waste and environmental problem. However, there is growing interest inmore » pyrolysis as a technology to treat tyres to produce valuable oil, char and gas products. The most common reactors used are fixed-bed (batch), screw kiln, rotary kiln, vacuum and fluidised-bed. The key influence on the product yield, and gas and oil composition, is the type of reactor used which in turn determines the temperature and heating rate. Tyre pyrolysis oil is chemically very complex containing aliphatic, aromatic, hetero-atom and polar fractions. The fuel characteristics of the tyre oil shows that it is similar to a gas oil or light fuel oil and has been successfully combusted in test furnaces and engines. The main gases produced from the pyrolysis of waste tyres are H{sub 2}, C{sub 1}–C{sub 4} hydrocarbons, CO{sub 2}, CO and H{sub 2}S. Upgrading tyre pyrolysis products to high value products has concentrated on char upgrading to higher quality carbon black and to activated carbon. The use of catalysts to upgrade the oil to a aromatic-rich chemical feedstock or the production of hydrogen from waste tyres has also been reported. Examples of commercial and semi-commercial scale tyre pyrolysis systems show that small scale batch reactors and continuous rotary kiln reactors have been developed to commercial scale.« less

Nuclear Rocket Technology Conference

NASA Technical Reports Server (NTRS)

1966-01-01

The Lewis Research Center has a strong interest in nuclear rocket propulsion and provides active support of the graphite reactor program in such nonnuclear areas as cryogenics, two-phase flow, propellant heating, fluid systems, heat transfer, nozzle cooling, nozzle design, pumps, turbines, and startup and control problems. A parallel effort has also been expended to evaluate the engineering feasibility of a nuclear rocket reactor using tungsten-matrix fuel elements and water as the moderator. Both of these efforts have resulted in significant contributions to nuclear rocket technology. Many successful static firings of nuclear rockets have been made with graphite-core reactors. Sufficient information has also been accumulated to permit a reasonable Judgment as to the feasibility of the tungsten water-moderated reactor concept. We therefore consider that this technoIogy conference on the nuclear rocket work that has been sponsored by the Lewis Research Center is timely. The conference has been prepared by NASA personnel, but the information presented includes substantial contributions from both NASA and AEC contractors. The conference excludes from consideration the many possible mission requirements for nuclear rockets. Also excluded is the direct comparison of nuclear rocket types with each other or with other modes of propulsion. The graphite reactor support work presented on the first day of the conference was partly inspired through a close cooperative effort between the Cleveland extension of the Space Nuclear Propulsion Office (headed by Robert W. Schroeder) and the Lewis Research Center. Much of this effort was supervised by Mr. John C. Sanders, chairman for the first day of the conference, and by Mr. Hugh M. Henneberry. The tungsten water-moderated reactor concept was initiated at Lewis by Mr. Frank E. Rom and his coworkers. The supervision of the recent engineering studies has been shared by Mr. Samuel J. Kaufman, chairman for the second day of the conference, and Mr. Roy V. Humble. Dr. John C. Eward served as general chairman for the conference.

High power ring methods and accelerator driven subcritical reactor application

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

Tahar, Malek Haj

2016-08-07

High power proton accelerators allow providing, by spallation reaction, the neutron fluxes necessary in the synthesis of fissile material, starting from Uranium 238 or Thorium 232. This is the basis of the concept of sub-critical operation of a reactor, for energy production or nuclear waste transmutation, with the objective of achieving cleaner, safer and more efficient process than today’s technologies allow. Designing, building and operating a proton accelerator in the 500-1000 MeV energy range, CW regime, MW power class still remains a challenge nowadays. There is a limited number of installations at present achieving beam characteristics in that class, e.g.,more » PSI in Villigen, 590 MeV CW beam from a cyclotron, SNS in Oakland, 1 GeV pulsed beam from a linear accelerator, in addition to projects as the ESS in Europe, a 5 MW beam from a linear accelerator. Furthermore, coupling an accelerator to a sub-critical nuclear reactor is a challenging proposition: some of the key issues/requirements are the design of a spallation target to withstand high power densities as well as ensure the safety of the installation. These two domains are the grounds of the PhD work: the focus is on the high power ring methods in the frame of the KURRI FFAG collaboration in Japan: upgrade of the installation towards high intensity is crucial to demonstrate the high beam power capability of FFAG. Thus, modeling of the beam dynamics and benchmarking of different codes was undertaken to validate the simulation results. Experimental results revealed some major losses that need to be understood and eventually overcome. By developing analytical models that account for the field defects, one identified major sources of imperfection in the design of scaling FFAG that explain the important tune variations resulting in the crossing of several betatron resonances. A new formula is derived to compute the tunes and properties established that characterize the effect of the field imperfections on the transverse beam dynamics. The results obtained allow to develop a correction scheme to minimize the tune variations of the FFAG. This is the cornerstone of a new fixed tune non-scaling FFAG that represents a potential candidate for high power applications. As part of the developments towards high power at the KURRI FFAG, beam dynamics studies have to account for space charge effects. In that framework, models have been installed in the tracking code ZGOUBI to account for the self-interaction of the particles in the accelerator. Application to the FFAG studies is shown. Finally, one focused on the ADSR concept as a candidate to solve the problem of nuclear waste. In order to establish the accelerator requirements, one compared the performance of ADSR with other conventional critical reactors by means of the levelized cost of energy. A general comparison between the different accelerator technologies that can satisfy these requirements is finally presented. In summary, the main drawback of the ADSR technology is the high Levelized Cost Of Energy compared to other advanced reactor concepts that do not employ an accelerator. Nowadays, this is a show-stopper for any industrial application aiming at producing energy (without dealing with the waste problem). Besides, the reactor is not intrinsically safer than critical reactor concepts, given the complexity of managing the target interface between the accelerator and the reactor core.« less

Reducing Actinide Production Using Inert Matrix Fuels

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

Deinert, Mark

2017-08-23

The environmental and geopolitical problems that surround nuclear power stem largely from the longlived transuranic isotopes of Am, Cm, Np and Pu that are contained in spent nuclear fuel. New methods for transmuting these elements into more benign forms are needed. Current research efforts focus largely on the development of fast burner reactors, because it has been shown that they could dramatically reduce the accumulation of transuranics. However, despite five decades of effort, fast reactors have yet to achieve industrial viability. A critical limitation to this, and other such strategies, is that they require a type of spent fuel reprocessingmore » that can efficiently separate all of the transuranics from the fission products with which they are mixed. Unfortunately, the technology for doing this on an industrial scale is still in development. In this project, we explore a strategy for transmutation that can be deployed using existing, current generation reactors and reprocessing systems. We show that use of an inert matrix fuel to recycle transuranics in a conventional pressurized water reactor could reduce overall production of these materials by an amount that is similar to what is achievable using proposed fast reactor cycles. Furthermore, we show that these transuranic reductions can be achieved even if the fission products are carried into the inert matrix fuel along with the transuranics, bypassing the critical separations hurdle described above. The implications of these findings are significant, because they imply that inert matrix fuel could be made directly from the material streams produced by the commercially available PUREX process. Zirconium dioxide would be an ideal choice of inert matrix in this context because it is known to form a stable solid solution with both fission products and transuranics.« less

Automated power control system for reactor TRIGA PUSPATI

NASA Astrophysics Data System (ADS)

Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

2017-01-01

Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

Modeling Lab-sized Anaerobic Fluidized Bed Reactor (AFBR) for Palm Oil Mill Effluent (POME) treatment: from Batch to Continuous Reactors

NASA Astrophysics Data System (ADS)

Mufti Azis, Muhammad; Sudibyo, Hanifrahmawan; Budhijanto, Wiratni

2018-03-01

Indonesia is aiming to produce 30 million tones/year of crude palm oil (CPO) by 2020. As a result, 90 million tones/year of POME will be produced. POME is highly polluting wastewater which may cause severe environmental problem due to its high chemical oxygen demand (COD) and biochemical oxygen demand (BOD). Due to the limitation of open pond treatment, the use of AFBR has been considered as a potential technology to treat POME. This study aims to develop mathematical models of lab-sized Anaerobic Fluidized Bed Reactor (AFBR) in batch and continuous processes. In addition, the AFBR also utilized natural zeolite as an immobilized media for microbes. To initiate the biomass growth, biodiesel waste has been used as an inoculum. In the first part of this study, a batch AFBR was operated to evaluate the COD, VFA, and CH4 concentrations. By comparing the batch results with and without zeolite, it showed that the addition of 17 g/gSCOD zeolite gave larger COD decrease within 20 days of operation. In order to elucidate the mechanism, parameter estimations of 12 kinetic parameters were proposed to describe the batch reactor performance. The model in general could describe the batch experimental data well. In the second part of this study, the kinetic parameters obtained from batch reactor were used to simulate the performance of double column AFBR where the acidogenic and methanogenic biomass were separated. The simulation showed that a relatively long residence time (Hydraulic Residence Time, HRT) was required to treat POME using the proposed double column AFBR. Sensitivity analyses was conducted and revealed that μm1 appeared to be the most sensitive parameter to reduce the HRT of double column AFBR.

Hybrid finite-volume/transported PDF method for the simulation of turbulent reactive flows

NASA Astrophysics Data System (ADS)

Raman, Venkatramanan

A novel computational scheme is formulated for simulating turbulent reactive flows in complex geometries with detailed chemical kinetics. A Probability Density Function (PDF) based method that handles the scalar transport equation is coupled with an existing Finite Volume (FV) Reynolds-Averaged Navier-Stokes (RANS) flow solver. The PDF formulation leads to closed chemical source terms and facilitates the use of detailed chemical mechanisms without approximations. The particle-based PDF scheme is modified to handle complex geometries and grid structures. Grid-independent particle evolution schemes that scale linearly with the problem size are implemented in the Monte-Carlo PDF solver. A novel algorithm, in situ adaptive tabulation (ISAT) is employed to ensure tractability of complex chemistry involving a multitude of species. Several non-reacting test cases are performed to ascertain the efficiency and accuracy of the method. Simulation results from a turbulent jet-diffusion flame case are compared against experimental data. The effect of micromixing model, turbulence model and reaction scheme on flame predictions are discussed extensively. Finally, the method is used to analyze the Dow Chlorination Reactor. Detailed kinetics involving 37 species and 158 reactions as well as a reduced form with 16 species and 21 reactions are used. The effect of inlet configuration on reactor behavior and product distribution is analyzed. Plant-scale reactors exhibit quenching phenomena that cannot be reproduced by conventional simulation methods. The FV-PDF method predicts quenching accurately and provides insight into the dynamics of the reactor near extinction. The accuracy of the fractional time-stepping technique in discussed in the context of apparent multiple-steady states observed in a non-premixed feed configuration of the chlorination reactor.

GAMSOR: Gamma Source Preparation and DIF3D Flux Solution

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

Smith, M. A.; Lee, C. H.; Hill, R. N.

2017-06-28

Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron capture reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problemsmore » with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, and then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence.« less

Thermodynamic analysis of in situ gasification-chemical looping combustion (iG-CLC) of Indian coal.

PubMed

Suresh, P V; Menon, Kavitha G; Prakash, K S; Prudhvi, S; Anudeep, A

2016-10-01

Chemical looping combustion (CLC) is an inherent CO 2 capture technology. It is gaining much interest in recent years mainly because of its potential in addressing climate change problems associated with CO 2 emissions from power plants. A typical chemical looping combustion unit consists of two reactors-fuel reactor, where oxidation of fuel occurs with the help of oxygen available in the form of metal oxides and, air reactor, where the reduced metal oxides are regenerated by the inflow of air. These oxides are then sent back to the fuel reactor and the cycle continues. The product gas from the fuel reactor contains a concentrated stream of CO 2 which can be readily stored in various forms or used for any other applications. This unique feature of inherent CO 2 capture makes the technology more promising to combat the global climate changes. Various types of CLC units have been discussed in literature depending on the type of fuel burnt. For solid fuel combustion three main varieties of CLC units exist namely: syngas CLC, in situ gasification-CLC (iG-CLC) and chemical looping with oxygen uncoupling (CLOU). In this paper, theoretical studies on the iG-CLC unit burning Indian coal are presented. Gibbs free energy minimization technique is employed to determine the composition of flue gas and oxygen carrier of an iG-CLC unit using Fe 2 O 3 , CuO, and mixed carrier-Fe 2 O 3 and CuO as oxygen carriers. The effect of temperature, suitability of oxygen carriers, and oxygen carrier circulation rate on the performance of a CLC unit for Indian coal are studied and presented. These results are analyzed in order to foresee the operating conditions at which economic and smooth operation of the unit is expected.

Absorption of Some Mineral Salts by Root System of Different Woody Species and Accumulation over a Whole Vegetative Cycle; ABSORPTION DE QUELQUES SELS PAR L'APPAREIL RADICULAIRE DE DIFFERENTES ESPECES LIGNEUSES ET ACCUMULATION AU COURS D'UN CYCLE VEGETATIF COMPLET

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

Gagnaire, J.

1963-01-01

The concentration power of plant tissues and the translocation speed of mineral salts vary considerably with the absorbed salt, the botanical species, the considered tissue, and the part of the vegetative cycle. In Grenoble, with Picea excelsa, the true dormance is short and is accompanied by a pre-dormance period and a post dormance period. In the vegetative period, Picea excelsa leaves concentrate less mineral salt than Acer campestris leaves (coefficient 2 for Ca--3 for phosphates) and Populus nigra leaves (coefficient 3 for Ca, coefficient 5 for phosphates). Results of tracer studies are tabulated. (C.H.)

[Case study of red water phenomenon in drinking water distribution systems caused by water source switch].

PubMed

Wang, Yang; Zhang, Xiao-jian; Chen, Chao; Pan, An-jun; Xu, Yang; Liao, Ping-an; Zhang, Su-xia; Gu, Jun-nong

2009-12-01

Red water phenomenon occurred in some communities of a city in China after water source switch in recent days. The origin of this red water problem and mechanism of iron release were investigated in the study. Water quality of local and new water sources was tested and tap water quality in suffered area had been monitored for 3 months since red water occurred. Interior corrosion scales on the pipe which was obtained from the suffered area were analyzed by XRD, SEM, and EDS. Corrosion rates of cast iron under the conditions of two source water were obtained by Annular Reactor. The influence of different source water on iron release was studied by pipe section reactor to simulate the distribution systems. The results indicated that large increase of sulfate concentration by water source shift was regarded as the cause of red water problem. The Larson ratio increased from about 0.4 to 1.7-1.9 and the red water problem happened in the taps of some urban communities just several days after the new water source was applied. The mechanism of iron release was concluded that the stable shell of scales in the pipes had been corrupted by this kind of high-sulfate-concentration source water and it was hard to recover soon spontaneously. The effect of sulfate on iron release of the old cast iron was more significant than its effect on enhancing iron corrosion. The rate of iron release increased with increasing Larson ratio, and the correlation of them was nonlinear on the old cast-iron. The problem remained quite a long time even if the water source re-shifted into the blended one with only small ratio of the new source and the Larson ratio reduced to about 0.6.

Subplane-based Control Rod Decusping Techniques for the 2D/1D Method in MPACT

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

Graham, Aaron M; Collins, Benjamin S; Downar, Thomas

2017-01-01

The MPACT transport code is being jointly developed by Oak Ridge National Laboratory and the University of Michigan to serve as the primary neutron transport code for the Virtual Environment for Reactor Applications Core Simulator. MPACT uses the 2D/1D method to solve the transport equation by decomposing the reactor model into a stack of 2D planes. A fine mesh flux distribution is calculated in each 2D plane using the Method of Characteristics (MOC), then the planes are coupled axially through a 1D NEM-Pmore » $$_3$$ calculation. This iterative calculation is then accelerated using the Coarse Mesh Finite Difference method. One problem that arises frequently when using the 2D/1D method is that of control rod cusping. This occurs when the tip of a control rod falls between the boundaries of an MOC plane, requiring that the rodded and unrodded regions be axially homogenized for the 2D MOC calculations. Performing a volume homogenization does not properly preserve the reaction rates, causing an error known as cusping. The most straightforward way of resolving this problem is by refining the axial mesh, but this can significantly increase the computational expense of the calculation. The other way of resolving the partially inserted rod is through the use of a decusping method. This paper presents new decusping methods implemented in MPACT that can dynamically correct the rod cusping behavior for a variety of problems.« less

Introduction to the IWA task group on biofilm modeling.

PubMed

Noguera, D R; Morgenroth, E

2004-01-01

An International Water Association (IWA) Task Group on Biofilm Modeling was created with the purpose of comparatively evaluating different biofilm modeling approaches. The task group developed three benchmark problems for this comparison, and used a diversity of modeling techniques that included analytical, pseudo-analytical, and numerical solutions to the biofilm problems. Models in one, two, and three dimensional domains were also compared. The first benchmark problem (BM1) described a monospecies biofilm growing in a completely mixed reactor environment and had the purpose of comparing the ability of the models to predict substrate fluxes and concentrations for a biofilm system of fixed total biomass and fixed biomass density. The second problem (BM2) represented a situation in which substrate mass transport by convection was influenced by the hydrodynamic conditions of the liquid in contact with the biofilm. The third problem (BM3) was designed to compare the ability of the models to simulate multispecies and multisubstrate biofilms. These three benchmark problems allowed identification of the specific advantages and disadvantages of each modeling approach. A detailed presentation of the comparative analyses for each problem is provided elsewhere in these proceedings.

Field Demonstration of a Novel Biotreatment Process for Perchlorate Reduction in Groundwater

DTIC Science & Technology

2010-06-01

biological reduction and/or reaction with ZVI, and arsenic hexavalent chromium and/or uranium by adsorption on corrosion products. • Simple rugged...problems and troubleshooting measures ................................... 22 5.2 Laboratory Evaluation of Porosity Decrease and Corrosion Products...reactor when it was dismantled showing the heavy deposits of iron corrosion products and quasi total loss of porosity. Figure 5.14 Picture of the column

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

Pickman, D.O.

Various aspects of zirconium alloy development for light water reactors in the UK and Scandinavia are reviewed, including the contribution made by some unique nuclear testing facilities. Among the problems encountered were the irradiation enhancement of corrosion and hydrogen pickup, crud deposition, iodine-induced stress-corrosion cracking on power ramping, and severe cladding deformation in loss-of-coolant accident conditions. The causes and behavior of defects, including hydride defects and fretting corrosion, are discussed.

ORIGEN-based Nuclear Fuel Inventory Module for Fuel Cycle Assessment: Final Project Report

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

Skutnik, Steven E.

The goal of this project, “ORIGEN-based Nuclear Fuel Depletion Module for Fuel Cycle Assessment" is to create a physics-based reactor depletion and decay module for the Cyclus nuclear fuel cycle simulator in order to assess nuclear fuel inventories over a broad space of reactor operating conditions. The overall goal of this approach is to facilitate evaluations of nuclear fuel inventories for a broad space of scenarios, including extended used nuclear fuel storage and cascading impacts on fuel cycle options such as actinide recovery in used nuclear fuel, particularly for multiple recycle scenarios. The advantages of a physics-based approach (compared tomore » a recipe-based approach which has been typically employed for fuel cycle simulators) is in its inherent flexibility; such an approach can more readily accommodate the broad space of potential isotopic vectors that may be encountered under advanced fuel cycle options. In order to develop this flexible reactor analysis capability, we are leveraging the Origen nuclear fuel depletion and decay module from SCALE to produce a standalone “depletion engine” which will serve as the kernel of a Cyclus-based reactor analysis module. The ORIGEN depletion module is a rigorously benchmarked and extensively validated tool for nuclear fuel analysis and thus its incorporation into the Cyclus framework can bring these capabilities to bear on the problem of evaluating long-term impacts of fuel cycle option choices on relevant metrics of interest, including materials inventories and availability (for multiple recycle scenarios), long-term waste management and repository impacts, etc. Developing this Origen-based analysis capability for Cyclus requires the refinement of the Origen analysis sequence to the point where it can reasonably be compiled as a standalone sequence outside of SCALE; i.e., wherein all of the computational aspects of Origen (including reactor cross-section library processing and interpolation, input and output processing, and depletion/decay solvers) can be self-contained into a single executable sequence. Further, to embed this capability into other software environments (such as the Cyclus fuel cycle simulator) requires that Origen’s capabilities be encapsulated into a portable, self-contained library which other codes can then call directly through function calls, thereby directly accessing the solver and data processing capabilities of Origen. Additional components relevant to this work include modernization of the reactor data libraries used by Origen for conducting nuclear fuel depletion calculations. This work has included the development of new fuel assembly lattices not previously available (such as for CANDU heavy-water reactor assemblies) as well as validation of updated lattices for light-water reactors updated to employ modern nuclear data evaluations. The CyBORG reactor analysis module as-developed under this workscope is fully capable of dynamic calculation of depleted fuel compositions from all commercial U.S. reactor assembly types as well as a number of international fuel types, including MOX, VVER, MAGNOX, and PHWR CANDU fuel assemblies. In addition, the Origen-based depletion engine allows for CyBORG to evaluate novel fuel assembly and reactor design types via creation of Origen reactor data libraries via SCALE. The establishment of this new modeling capability affords fuel cycle modelers a substantially improved ability to model dynamically-changing fuel cycle and reactor conditions, including recycled fuel compositions from fuel cycle scenarios involving material recycle into thermal-spectrum systems.« less

Isolation of Metals from Liquid Wastes: Reactive Scavenging in Turbulent Thermal Reactors

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

Jost O.L. Wendt; Alan R. Kerstein; Alexander Scheeline

2003-08-06

The Overall project demonstrated that toxic metals (cesium Cs and strontium Sr) in aqueous and organic wastes can be isolated from the environment through reaction with kaolinite based sorbent substrates in high temperature reactor environments. In addition, a state-of-the art laser diagnostic tool to measure droplet characteristic in practical 'dirty' laboratory environments was developed, and was featured on the cover of a recent edition of the scientific journal ''applied Spectroscopy''. Furthermore, great strides have been made in developing a theoretical model that has the potential to allow prediction of the position and life history of every particle of waste inmore » a high temperature, turbulent flow field, a very challenging problem involving as it does, the fundamentals of two phase turbulence and of particle drag physics.« less

Foam suppression in overloaded manure-based biogas reactors using antifoaming agents.

PubMed

Kougias, P G; Boe, K; Tsapekos, P; Angelidaki, I

2014-02-01

Foam control is an imperative need in biogas plants, as foaming is a major operational problem. In the present study, the effect of oils (rapeseed oil, oleic acid, and octanoic acid) and tributylphosphate on foam reduction and process performance in batch and continuous manure-based biogas reactors was investigated. The compounds were tested in dosages of 0.05%, 0.1% and 0.5% v/vfeed. The results showed that rapeseed oil was most efficient to suppress foam at the dosage of 0.05% and 0.1% v/vfeed, while octanoic acid was most efficient to suppress foam at dosage of 0.5% v/vfeed. Moreover, the addition of rapeseed oil also increased methane yield. In contrast, tributylphosphate, which was very efficient antifoam, was found to be inhibitory to the biogas process. Copyright © 2013 Elsevier Ltd. All rights reserved.

NASA-Lewis experiences with multigroup cross sections and shielding calculations

NASA Technical Reports Server (NTRS)

Lahti, G. P.

1972-01-01

The nuclear reactor shield analysis procedures employed at NASA-Lewis are described. Emphasis is placed on the generation, use, and testing of multigroup cross section data. Although coupled neutron and gamma ray cross section sets are useful in two dimensional Sn transport calculations, much insight has been gained from examination of uncoupled calculations. These have led to experimental and analytic studies of areas deemed to be of first order importance to reactor shield calculations. A discussion is given of problems encountered in using multigroup cross sections in the resolved resonance energy range. The addition to ENDF files of calculated and/or measured neutron-energy-dependent capture gamma ray spectra for shielding calculations is questioned for the resonance region. Anomalies inherent in two dimensional Sn transport calculations which may overwhelm any cross section discrepancies are illustrated.

Multiscale Multiphysics Developments for Accident Tolerant Fuel Concepts

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

Gamble, K. A.; Hales, J. D.; Yu, J.

2015-09-01

U 3Si 2 and iron-chromium-aluminum (Fe-Cr-Al) alloys are two of many proposed accident-tolerant fuel concepts for the fuel and cladding, respectively. The behavior of these materials under normal operating and accident reactor conditions is not well known. As part of the Department of Energy’s Accident Tolerant Fuel High Impact Problem program significant work has been conducted to investigate the U 3Si 2 and FeCrAl behavior under reactor conditions. This report presents the multiscale and multiphysics effort completed in fiscal year 2015. The report is split into four major categories including Density Functional Theory Developments, Molecular Dynamics Developments, Mesoscale Developments, andmore » Engineering Scale Developments. The work shown here is a compilation of a collaborative effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory and Anatech Corp.« less

Compared microbiology of granular sludge under autotrophic, mixotrophic and heterotrophic denitrification conditions.

PubMed

Fernández, N; Sierra-Alvarez, R; Amils, R; Field, J A; Sanz, J L

2009-01-01

Water contamination by nitrate is a wideworld extended phenomena. Biological autotrophic denitrification has a real potential to face this problem and presents less drawbacks than the most extended heterotrophic denitrification. Three bench-scale UASB reactors were operated under autotrophic (R1, H2S as electron donor), mixotrophic (R2, H2S plus p-cresol as electron donors) and heterotrophic (R3, p-cresol as electron donor) conditions using nitrate as terminal electron acceptor. 16S rDNA genetic libraries were built up to compare their microbial biodiversity. Six different bacteria phyla and three archaeal classes were observed. Proteobacteria was the main phyla in all reactors standing out the presence of denitrifiers. Microorganisms similar to Thiobacillus denitrificans and Acidovorax sp. performed the autotrophic denitification. These OTUs were displaced by chemoheterotrophic denitrifiers, especially by Limnobacter-like and Ottowia-like OTUs. Other phyla were Bacteroidetes, Chloroflexi, Firmicutes and Actinobacteria that--as well as Archaea members--were implicated in the degradation of organic matter, as substrate added as coming from endogenous sludge decay under autotrophic conditions. Archaea diversity remained low in all the reactors being Methanosaeta concilii the most abundant one.

Long-term proliferation and safeguards issues in future technologies

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

Keisch, B.; Auerbach, C.; Fainberg, A.

1986-02-01

The purpose of the task was to assess the effect of potential new technologies, nuclear and non-nuclear, on safeguards needs and non-proliferation policies, and to explore possible solutions to some of the problems envisaged. Eight subdivisions were considered: New Enrichment Technologies; Non-Aqueous Reprocessing Technologies; Fusion; Accelerator-Driven Reactor Systems; New Reactor Types; Heavy Water and Deuterium; Long-Term Storage of Spent Fuel; and Other Future Technologies (Non-Nuclear). For each of these subdivisions, a careful review of the current world-wide effort in the field provided a means of subjectively estimating the viability and qualitative probability of fruition of promising technologies. Technologies for whichmore » safeguards and non-proliferation requirements have been thoroughly considered by others were not restudied here (e.g., the Fast Breeder Reactor). The time scale considered was 5 to 40 years for possible initial demonstration although, in some cases, a somewhat optimistic viewpoint was embraced. Conventional nuclear-material safeguards are only part of the overall non-proliferation regime. Other aspects are international agreements, export controls on sensitive technologies, classification of information, intelligence gathering, and diplomatic initiatives. The focus here is on safeguards, export controls, and classification.« less

Structure and activity of a new low-molecular-weight heparin produced by enzymatic ultrafiltration.

PubMed

Fu, Li; Zhang, Fuming; Li, Guoyun; Onishi, Akihiro; Bhaskar, Ujjwal; Sun, Peilong; Linhardt, Robert J

2014-05-01

The standard process for preparing the low-molecular-weight heparin (LMWH) tinzaparin, through the partial enzymatic depolymerization of heparin, results in a reduced yield because of the formation of a high content of undesired disaccharides and tetrasaccharides. An enzymatic ultrafiltration reactor for LMWH preparation was developed to overcome this problem. The behavior, of the heparin oligosaccharides and polysaccharides using various membranes and conditions, was investigated to optimize this reactor. A novel product, LMWH-II, was produced from the controlled depolymerization of heparin using heparin lyase II in this optimized ultrafiltration reactor. Enzymatic ultrafiltration provides easy control and high yields (>80%) of LMWH-II. The molecular weight properties of LMWH-II were similar to other commercial LMWHs. The structure of LMWH-II closely matched heparin's core structural features. Most of the common process artifacts, present in many commercial LWMHs, were eliminated as demonstrated by 1D and 2D nuclear magnetic resonance spectroscopy. The antithrombin III and platelet factor-4 binding affinity of LMWH-II were comparable to commercial LMWHs, as was its in vitro anticoagulant activity. © 2014 Wiley Periodicals, Inc. and the American Pharmacists Association.

Optimization of operating parameters in polysilicon chemical vapor deposition reactor with response surface methodology

NASA Astrophysics Data System (ADS)

An, Li-sha; Liu, Chun-jiao; Liu, Ying-wen

2018-05-01

In the polysilicon chemical vapor deposition reactor, the operating parameters are complex to affect the polysilicon's output. Therefore, it is very important to address the coupling problem of multiple parameters and solve the optimization in a computationally efficient manner. Here, we adopted Response Surface Methodology (RSM) to analyze the complex coupling effects of different operating parameters on silicon deposition rate (R) and further achieve effective optimization of the silicon CVD system. Based on finite numerical experiments, an accurate RSM regression model is obtained and applied to predict the R with different operating parameters, including temperature (T), pressure (P), inlet velocity (V), and inlet mole fraction of H2 (M). The analysis of variance is conducted to describe the rationality of regression model and examine the statistical significance of each factor. Consequently, the optimum combination of operating parameters for the silicon CVD reactor is: T = 1400 K, P = 3.82 atm, V = 3.41 m/s, M = 0.91. The validation tests and optimum solution show that the results are in good agreement with those from CFD model and the deviations of the predicted values are less than 4.19%. This work provides a theoretical guidance to operate the polysilicon CVD process.

Valorisation of waste tyre by pyrolysis in a moving bed reactor.

PubMed

Aylón, E; Fernández-Colino, A; Murillo, R; Navarro, M V; García, T; Mastral, A M

2010-07-01

The aim of this work is to assess the behaviour of a moving bed reactor, based on a screw transporter design, in waste tyre pyrolysis under several experimental conditions. Waste tyre represents a significant problem in developed countries and it is necessary to develop new technology that could easily process big amounts of this potentially raw material. In this work, the influence of the main pyrolysis process variables (temperature, solid residence time, mass flow rate and inert gas flow) has been studied by a thorough analysis of product yields and properties. It has been found that regardless the process operational parameters, a total waste tyre devolatilisation is achieved, producing a pyrolytic carbon black with a volatile matter content under 5 wt.%. In addition, it has been proven that, in the range studied, the most influencing process variables are temperature and solid mass flow rate, mainly because both variables modify the gas residence time inside the reactor. In addition, it has been found that the modification of these variables affects to the chemical properties of the products. This fact is mainly associated to the different cracking reaction of the primary pyrolysis products. Copyright (c) 2009 Elsevier Ltd. All rights reserved.

Valorisation of waste tyre by pyrolysis in a moving bed reactor

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

Aylon, E.; Fernandez-Colino, A.; Murillo, R., E-mail: ramonm@icb.csic.e

2010-07-15

The aim of this work is to assess the behaviour of a moving bed reactor, based on a screw transporter design, in waste tyre pyrolysis under several experimental conditions. Waste tyre represents a significant problem in developed countries and it is necessary to develop new technology that could easily process big amounts of this potentially raw material. In this work, the influence of the main pyrolysis process variables (temperature, solid residence time, mass flow rate and inert gas flow) has been studied by a thorough analysis of product yields and properties. It has been found that regardless the process operationalmore » parameters, a total waste tyre devolatilisation is achieved, producing a pyrolytic carbon black with a volatile matter content under 5 wt.%. In addition, it has been proven that, in the range studied, the most influencing process variables are temperature and solid mass flow rate, mainly because both variables modify the gas residence time inside the reactor. In addition, it has been found that the modification of these variables affects to the chemical properties of the products. This fact is mainly associated to the different cracking reaction of the primary pyrolysis products.« less

Dose Rate Calculation of TRU Metal Ingot in Pyroprocessing - 12202

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

Lee, Yoon Hee; Lee, Kunjai

Spent fuel management has been a main problem to be solved for continuous utilization of nuclear energy. Spent fuel management policy of Korea is 'Wait and See'. It is focused on Pyro-process and SFR (Sodium-cooled Fast Reactor) for closed-fuel cycle research and development in Korea. For peaceful use of nuclear facilities, the proliferation resistance has to be proved. Proliferation resistance is one of key constraints in the deployment of advanced nuclear energy systems. Non-proliferation and safeguard issues have been strengthening internationally. Barriers to proliferation are that reduces desirability or attractiveness as an explosive and makes it difficult to gain accessmore » to the materials, or makes it difficult to misuse facilities and/or technologies for weapons applications. Barriers to proliferation are classified into intrinsic and extrinsic barriers. Intrinsic barrier is inherent quality of reactor materials or the fuel cycle that is built into the reactor design and operation such as material and technical barriers. As one of the intrinsic measures, the radiation from the material is considered significantly. Therefore the radiation of TRU metal ingot from the pyro-process was calculated using ORIGEN and MCNP code. (authors)« less

Efficient removal of antibiotics in a fluidized bed reactor by facile fabricated magnetic powdered activated carbon.

PubMed

Ma, Jianqing; Yang, Qunfeng; Xu, Dongmei; Zeng, Xiaomei; Wen, Yuezhong; Liu, Weiping

2017-02-01

Powdered activated carbons (PACs) with micrometer size are showing great potential for enabling and improving technologies in water treatment. The critical problem in achieving practical application of PAC involves simple, effective fabrication of magnetic PAC and the design of a feasible reactor that can remove pollutants and recover the adsorbent efficiently. Herein, we show that such materials can be fabricated by the combination of PAC and magnetic Fe 3 O 4 with chitosan-Fe hydrogel through a simple co-precipitation method. According to the characterization results, CS-Fe/Fe 3 O 4 /PAC with different micrometers in size exhibited excellent magnetic properties. The adsorption of tetracycline was fast and efficient, and 99.9% removal was achieved in 30 min. It also possesses good usability and stability to co-existing ions, organics, and different pH values due to its dispersive interaction nature. Finally, the prepared CS-Fe/Fe 3 O 4 /PAC also performed well in the fluidized bed reactor with electromagnetic separation function. It could be easily separated by applying a magnetic field and was effectively in situ regenerated, indicating a potential of practical application for the removal of pollutants from water.

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

NASA Astrophysics Data System (ADS)

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

2018-02-01

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

Effect of sludge retention time on the biological performance of anaerobic membrane bioreactors treating corn-to-ethanol thin stillage with high lipid content.

PubMed

Dereli, Recep Kaan; van der Zee, Frank P; Heffernan, Barry; Grelot, Aurelie; van Lier, Jules B

2014-02-01

The potential of anaerobic membrane bioreactors (AnMBRs) for the treatment of lipid rich corn-to-ethanol thin stillage was investigated at three different sludge retention times (SRT), i.e. 20, 30 and 50 days. The membrane assisted biomass retention in AnMBRs provided an excellent solution to sludge washout problems reported for the treatment of lipid rich wastewaters by granular sludge bed reactors. The AnMBRs achieved high COD removal efficiencies up to 99% and excellent effluent quality. Although higher organic loading rates (OLRs) up to 8.0 kg COD m(-3) d(-1) could be applied to the reactors operated at shorter SRTs, better biological degradation efficiencies, i.e. up to 83%, was achieved at increased SRTs. Severe long chain fatty acid (LCFA) inhibition was observed at 50 days SRT, possibly caused by the extensive dissolution of LCFA in the reactor broth, inhibiting the methanogenic biomass. Physicochemical mechanisms such as precipitation with divalent cations and adsorption on the sludge played an important role in the occurrence of LCFA removal, conversion, and inhibition. Copyright © 2013 Elsevier Ltd. All rights reserved.

Oily wastewater treatment using a novel hybrid PBR-UASB system.

PubMed

Jeganathan, Jeganaesan; Nakhla, George; Bassi, Amarjeet

2007-04-01

In this study, anaerobic treatability of oily wastewater was investigated in a hybrid reactor system consisting of a packed bed reactor (PBR) followed by an upflow anaerobic sludge blanket (UASB) reactor at 35 degrees C. The system was operated using real pet food wastewater at different hydraulic retention times and loading rates for 165 d. The PBR was packed with sol-gel/alginate beads containing immobilized enzyme which hydrolyzed the oil and grease (O&G) into free long chain fatty acids, that were biodegraded by the UASB. The hybrid system was operated up to an oil loading rate of 4.9 kg O&Gm(-3)d(-1) (to the PBR) without any operational problems for a period of 100 d, with COD and O&G removal efficiencies above 90% and no sludge flotation was observed in the UASB. Beads supplement to the PBR was less than 2 g d(-1) and the relative activity was about 70%. Further increment in O&G loading to 18.7 kg O&Gm(-3)d(-1) caused destabilization of the system with 0.35% (v float/v feed) sludge float removed from the UASB.

Evolution of a phase separated gravity independent bioreactor

NASA Technical Reports Server (NTRS)

Villeneuve, Peter E.; Dunlop, Eric H.

1992-01-01

The evolution of a phase-separated gravity-independent bioreactor is described. The initial prototype, a zero head-space manifold silicone membrane based reactor, maintained large diffusional resistances. Obtaining oxygen transfer rates needed to support carbon-recycling aerobic microbes is impossible if large resistances are maintained. Next generation designs (Mark I and II) mimic heat exchanger design to promote turbulence at the tubing-liquid interface, thereby reducing liquid and gas side diffusional resistances. While oxygen transfer rates increased by a factor of ten, liquid channeling prevented further increases. To overcome these problems, a Mark III reactor was developed which maintains inverted phases, i.e., media flows inside the silicone tubing, oxygen gas is applied external to the tubing. This enhances design through changes in gas side driving force concentration and liquid side turbulence levels. Combining an applied external pressure of 4 atm with increased Reynolds numbers resulted in oxygen transfer intensities of 232 mmol O2/l per hr (1000 times greater than the first prototype and comparable to a conventional fermenter). A 1.0 liter Mark III reactor can potentially deliver oxygen supplies necessary to support cell cultures needed to recycle a 10-astronaut carbon load continuously.

In situ bioremediation in Europe

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

Porta, A.; Young, J.K.; Molton, P.M.

1993-06-01

Site remediation activity in Europe is increasing, even if not at the forced pace of the US. Although there is a better understanding of the benefits of bioremediation than of other approaches, especially about in situ bioremediation of contaminated soils, relatively few projects have been carried out full-scale in Europe or in the US. Some engineering companies and large industrial companies in Europe are investigating bioremediation and biotreatment technologies, in some cases to solve their internal waste problems. Technologies related to the application of microorganisms to the soil, release of nutrients into the soil, and enhancement of microbial decontamination aremore » being tested through various additives such as surfactants, ion exchange resins, limestone, or dolomite. New equipment has been developed for crushing and mixing or injecting and sparging the microorganisms, as have new reactor technologies (e.g., rotating aerator reactors, biometal sludge reactors, and special mobile containers for simultaneous storage, transportation, and biodegradation of contaminated soil). Some work has also been done with immobilized enzymes to support and restore enzymatic activities related to partial or total xenobiotic decontamination. Finally, some major programs funded by public and private institutions confirm that increasing numbers of firms have a working interest in bioremediation.« less

THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

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

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures inmore » the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.« less

Pyrolysis of waste tyres: a review.

PubMed

Williams, Paul T

2013-08-01

Approximately 1.5 billion tyres are produced each year which will eventually enter the waste stream representing a major potential waste and environmental problem. However, there is growing interest in pyrolysis as a technology to treat tyres to produce valuable oil, char and gas products. The most common reactors used are fixed-bed (batch), screw kiln, rotary kiln, vacuum and fluidised-bed. The key influence on the product yield, and gas and oil composition, is the type of reactor used which in turn determines the temperature and heating rate. Tyre pyrolysis oil is chemically very complex containing aliphatic, aromatic, hetero-atom and polar fractions. The fuel characteristics of the tyre oil shows that it is similar to a gas oil or light fuel oil and has been successfully combusted in test furnaces and engines. The main gases produced from the pyrolysis of waste tyres are H(2), C(1)-C(4) hydrocarbons, CO(2), CO and H(2)S. Upgrading tyre pyrolysis products to high value products has concentrated on char upgrading to higher quality carbon black and to activated carbon. The use of catalysts to upgrade the oil to a aromatic-rich chemical feedstock or the production of hydrogen from waste tyres has also been reported. Examples of commercial and semi-commercial scale tyre pyrolysis systems show that small scale batch reactors and continuous rotary kiln reactors have been developed to commercial scale. Copyright © 2013 Elsevier Ltd. All rights reserved.

Parents of two-phase flow and theory of "gas-lift"

NASA Astrophysics Data System (ADS)

Zitek, Pavel; Valenta, Vaclav

2014-03-01

This paper gives a brief overview of types of two-phase flow. Subsequently, it deals with their mutual division and problems with accuracy boundaries among particular types. It also shows the case of water flow through a pipe with external heating and the gradual origination of all kinds of flow. We have met it in solution of safety condition of various stages in pressurized and boiling water reactors. In the MSR there is a problem in the solution of gas-lift using helium as a gas and its secondary usage for clearing of the fuel mixture from gaseous fission products. Theory of gas-lift is described.

Prevalence of behavior problems in three-year-old children living near Three Mile Island: a comparative analysis.

PubMed

Cornely, P; Bromet, E

1986-07-01

The Behavior Screening Questionnaire (BSQ) was used to determine whether 2 1/2-3 1/2 yr old children living near the TMI nuclear reactor were more disturbed than children living near another nuclear plant or near a fossil-fuel facility in Pennsylvania when assessed 2 1/2 yr later. The prevalence of behavior problems was 11%. Differences among the sites in overall rates and individual symptoms were small. Perceptions of environmental stress among the TMI sample of mothers were unrelated to BSQ scores, whereas in the comparison sites, where unemployment was rising, economic concerns were meaningfully related to the BSQ.

Orbital Space Solar Power Option for a Lunar Village

NASA Technical Reports Server (NTRS)

Johnson, Les

2017-01-01

One of the most significant challenges to the implementation of a continuously manned lunar base is power. During the lunar day (14 Earth days), it is conceptually simple to deploy solar arrays to generate the estimated 35 kilowatts of continuous power required. However, generating this level of power during the lunar night (also 14 Earth days) has been an extremely difficult problem to solve. Conventional solutions range from the requirement that the base be located at the lunar south pole so as to take advantage of the continuous sunshine available there to developing a space-qualified nuclear reactor and power plant to generate the needed energy. There is a third option: Use the soon-to-be-available Space Launch System to place a space based solar power station in lunar orbit that would beam the needed energy to the lunar base. Several detailed studies have been performed by NASA, universities and others looking at the lunar south pole for locating the base. The results are encouraging: by taking advantage of the moon's orbital tilt, large solar arrays can be deployed there to track the sun continuously and generate the power needed to sustain the base. The problem with this approach is inherent to its design: it will only work at the lunar south pole. There is no other site on the Moon with geometry favorable to generating continuous solar power. NASA has also considered the development of a compact fission reactor and power plant to generate the needed power, allowing the base to be sited anywhere on the Moon. The problem with this approach is that there are no space fission reactors available, none are being planned and the cost of developing one is prohibitively expensive. Using an orbiting space based solar power station to generate electrical power and beam it to a base sited anywhere on the moon should therefore be considered. The technology to collect sunlight, generate greater than the estimated 35 kilowatts of power, and beam it to the surface using microwaves is available today. The problem with this concept in the past would have been the mass and packaging volume (for launch) required to put such a system in place in lunar orbit. This problem is potentially solved with the advent of the Space Launch System (SLS). The SLS, with its 70 mT launch capacity, it more than capable of placing such a system into lunar orbit in a single launch. This paper will examine the potential use of an SLS-launched, space solar power system in lunar orbit as the primary power source for a first-generation, continuously-occupied lunar base and compare it with the other power generation and storage options previously considered.

Control of H2S emissions using an ozone oxidation process: Preliminary results

NASA Technical Reports Server (NTRS)

Defaveri, D.; Ferrando, B.; Ferraiolo, G.

1986-01-01

The problem of eliminating industrial emission odors does not have a simple solution, and consequently has not been researched extensively. Therefore, an experimental research program regarding oxidation of H2S through ozone was undertaken to verify the applicable limits of the procedure and, in addition, was designed to supply a useful analytical means of rationalizing the design of reactors employed in the sector.

Mesoporous Aluminosilicates as a Host and Reactor for Preparation of Ordered Metal Nanowires

NASA Astrophysics Data System (ADS)

Eliseev, A. A.; Napolskii, K. S.; Kolesnik, I. V.; Kolenko, Yu. V.; Lukashin, A. V.; Gornert, P.; Tretyakov, Yu. D.

The creation of functional nanomaterials with the controlled properties is emerging as a new area of great technological and scientific interest, in particular, it is a key technology for developing novel high-density data storage devices. Today, no other technology can compete with magnetic carriers in information storage density and access rate. However, usually very small (10-1000 nm3) magnetic nanoparticles shows para- or superparamagnetic properties, with very low blocking temperatures and no coercitivity at normal conditions. One possible solution of this problem is preparation of highly anisotropic nanostructures. From the other hand, the use of purely nanocrystalline systems is limited because of their low stability and tendency to form aggregates. These problems could be solved by encapsulation of nanoparticles to a chemically inert matrix. One of the promising matrices for preparation of highly anisotropic magnetic nanoparticles is mesoporous silica or mesoporous aluminosilicates. Mesoporous silica is an amorphous SiO2 with a highly ordered uniform pore structure (the pore diameter can be controllably varied from 2 to 50 nm). This pore system is a perfect reactor for synthesis of nanocomposites due to the limitation of reaction zone by the pore walls. One could expect that size and shape of nanoparticles incorporated into mesoporous silica to be consistent with the dimensions of the porous framework.

Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

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

Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.

2008-06-23

This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been mademore » at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.« less

ATTRITION RESISTANT IRON-BASED FISCHER-TROPSCH CATALYSTS

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

JAMES G. GOODWIN, JR.; JAMES J. SPIVEY; K. JOTHIMURUGESAN

1998-09-17

The Fischer-Tropsch (F-T) reaction provides a way of converting coal-derived synthesis gas (CO+H{sub 2}) to liquid fuels. Since the reaction is highly exothermic, one of the major problems in control of the reaction is heat removal. Recent work has shown that the use of slurry bubble column reactors (SBCRs) can largely solve this problem. Iron-based (Fe) catalysts are preferred catalysts for F-T when using low CO/H{sub 2} ratio synthesis gases derived from modern coal gasifiers. This is because in addition to reasonable F-T activity, the F-T catalysts also possess high water gas shift (WGS) activity. However, a serious problem withmore » the use of Fe catalysts in a SBCR is their tendency to undergo attrition. This can cause fouling/plugging of downstream filters and equipment, making the separation of catalyst from the oil/wax product very difficult if not impossible, and results in a steady loss of catalyst from the reactor. The objectives of this research are to develop a better understanding of the parameters affecting attrition resistance of Fe F-T catalysts suitable for use in SBCRs and to incorporate this understanding into the design of novel Fe catalysts having superior attrition resistance. Catalyst preparations will be based on the use of spray drying and will be scalable using commercially available equipment. The research will employ among other measurements, attrition testing and F-T synthesis, including long duration slurry reactor runs in order to ascertain the degree of success of the various preparations. The goal is to develop an Fe catalyst which can be used in a SBCR having only an internal filter for separation of the catalyst from the liquid product, without sacrificing F-T activity and selectivity. The effect of silica addition via coprecipitation and as a binder to a doubly promoted Fischer-Tropsch synthesis iron catalyst (100 Fe/5 Cu/4.2 K) was studied. The catalysts were prepared by coprecipitation, followed by binder addition and drying in a 1 m diameter, 2 m tall spray dryer. The binder silica content was varied from 0 to 20 wt %. A catalyst with 12 wt % binder silica was found to have the highest attrition resistance. F-T reaction studies over 100 hours in a fixed-bed reactor showed that this catalyst maintained around 95 % CO conversion with a methane selectivity of less than 7 wt % and a C{sub 5}{sup +} selectivity of greater than 73 wt %. The effect of adding precipitated silica from 0 to 20 parts by weight to this catalyst (containing 12 wt % binder silica) was also studied. Addition of precipitated silica was found to be detrimental to attrition resistance and resulted in increased methane and reduced wax formation.« less

Andrei Sakharov

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

None

2006-07-07

Ce discours donné par Mons.Jonauch qui est né en Tchécoslovaquie et a fait ses études à Leningrad, Moscou et Prague, est organisé par le comité Youri Orlov. Le conférencier parle de Andrei Sakharov, ce physicien et homme soviétique qui fit ses études à Moscou, effectua des recherches sur les armes thermonucléaires et entra à l'Académie des Sciences d'URSS en 1953. Il participa à la mise au point de la bombe à hydrogène, mais s'opposa quelques années plus tard à la poursuite des expériences nucléaires. Il créa en 1970 le comité pour la défense des droits de l'homme ce que luimore » valut le prix Nobel de la paix en 1975.« less

Conceptual study of an ICRH traveling-wave antenna system for low-coupling conditions as expected in DEMO

NASA Astrophysics Data System (ADS)

Ragona, R.; Messiaen, A.

2016-07-01

For the central heating of a fusion reactor ion cyclotron radio frequency heating (ICRH) is the first choice method as it is able to couple RF power to the ions without density limit. The drawback of this heating method is the problem of excitation of the magneto-sonic wave through the plasma boundary layer from the antenna located along the wall, without exceeding its voltage standoff. The amount of coupling depends on the antenna excitation and the surface admittance at the antenna output due to the plasma profile. The paper deals with the optimization of the antenna excitation by the use of sections of traveling-wave antennas (TWAs) distributed all along the reactor wall between the blanket modules. They are mounted and fed in resonant ring system(s). First, the physics of the coupling of a strap array is studied by simple models and the coupling code ANTITER II. Then, after the study of the basic properties of a TWA section, its feeding problem is solved by hybrids driving them in resonant ring circuit(s). The complete modeling is obtained from the matrices of the TWA sections connected to one of the feeding hybrid(s). The solution is iterated with the coupling code to determine the loading for a reference low-coupling ITER plasma profile. The resulting wave pattern up to the plasma bulk is derived. The proposed system is totally load resilient and allows us to obtain a very selective exciting wave spectrum. A discussion of some practical implementation problems is added.

Hairy root culture in a liquid-dispersed bioreactor: characterization of spatial heterogeneity.

PubMed

Williams, G R; Doran, P M

2000-01-01

A liquid-dispersed reactor equipped with a vertical mesh cylinder for inoculum support was developed for culture of Atropa belladonna hairy roots. The working volume of the culture vessel was 4.4 L with an aspect ratio of 1.7. Medium was dispersed as a spray onto the top of the root bed, and the roots grew radially outward from the central mesh cylinder to the vessel wall. Significant benefits in terms of liquid drainage and reduced interstitial liquid holdup were obtained using a vertical rather than horizontal support structure for the biomass and by operating the reactor with cocurrent air and liquid flow. With root growth, a pattern of spatial heterogeneity developed in the vessel. Higher local biomass densities, lower volumes of interstitial liquid, lower sugar concentrations, and higher root atropine contents were found in the upper sections of the root bed compared with the lower sections, suggesting a greater level of metabolic activity toward the top of the reactor. Although gas-liquid oxygen transfer to the spray droplets was very rapid, there was evidence of significant oxygen limitations in the reactor. Substantial volumes of non-free-draining interstitial liquid accumulated in the root bed. Roots near the bottom of the vessel trapped up to 3-4 times their own weight in liquid, thus eliminating the advantages of improved contact with the gas phase offered by liquid-dispersed culture systems. Local nutrient and product concentrations in the non-free-draining liquid were significantly different from those in the bulk medium, indicating poor liquid mixing within the root bed. Oxygen enrichment of the gas phase improved neither growth nor atropine production, highlighting the greater importance of liquid-solid compared with gas-liquid oxygen transfer resistance. The absence of mechanical or pneumatic agitation and the tendency of the root bed to accumulate liquid and impede drainage were identified as the major limitations to reactor performance. Improved reactor operating strategies and selection or development of root lines offering minimal resistance to liquid flow and low liquid retention characteristics are possible solutions to these problems.

Passive Acoustic Leak Detection for Sodium Cooled Fast Reactors Using Hidden Markov Models

NASA Astrophysics Data System (ADS)

Marklund, A. Riber; Kishore, S.; Prakash, V.; Rajan, K. K.; Michel, F.

2016-06-01

Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970s and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), the proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control.

Nuclear fission: the interplay of science and technology.

PubMed

Stoneham, A M

2010-07-28

When the UK's Calder Hall nuclear power station was connected to the grid in 1956, the programmes that made this possible involved a powerful combination of basic and applied research. Both the science and the engineering were novel, addressing new and challenging problems. That the last Calder Hall reactor was shut down only in 2003 attests to the success of the work. The strengths of bringing basic science to bear on applications continued to be recognized until the 1980s, when government and management fashions changed. This paper identifies a few of the technology challenges, and shows how novel basic science emerged from them and proved essential in their resolution. Today, as the threat of climate change becomes accepted, it has become clear that there is no credible solution without nuclear energy. The design and construction of new fission reactors will need continuing innovation, with the interplay between the science and technology being a crucial component.

Nonlinear versus Ordinary Adaptive Control of Continuous Stirred-Tank Reactor

PubMed Central

Dostal, Petr

2015-01-01

Unfortunately, the major group of the systems in industry has nonlinear behavior and control of such processes with conventional control approaches with fixed parameters causes problems and suboptimal or unstable control results. An adaptive control is one way to how we can cope with nonlinearity of the system. This contribution compares classic adaptive control and its modification with Wiener system. This configuration divides nonlinear controller into the dynamic linear part and the static nonlinear part. The dynamic linear part is constructed with the use of polynomial synthesis together with the pole-placement method and the spectral factorization. The static nonlinear part uses static analysis of the controlled plant for introducing the mathematical nonlinear description of the relation between the controlled output and the change of the control input. Proposed controller is tested by the simulations on the mathematical model of the continuous stirred-tank reactor with cooling in the jacket as a typical nonlinear system. PMID:26346878

The role of accelerators in the nuclear fuel cycle

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

Takahashi, Hiroshi.

1990-01-01

The use of neutrons produced by the medium energy proton accelerator (1 GeV--3 GeV) has considerable potential in reconstructing the nuclear fuel cycle. About 1.5 {approximately} 2.5 ton of fissile material can be produced annually by injecting a 450 MW proton beam directly into fertile materials. A source of neutrons, produced by a proton beam, to supply subcritical reactors could alleviate many of the safety problems associated with critical assemblies, such as positive reactivity coefficients due to coolant voiding. The transient power of the target can be swiftly controlled by controlling the power of the proton beam. Also, the usemore » of a proton beam would allow more flexibility in the choice of fuel and structural materials which otherwise might reduce the reactivity of reactors. This paper discusses the rate of accelerators in the transmutation of radioactive wastes of the nuclear fuel cycles. 34 refs., 17 figs., 9 tabs.« less

Electrochemical processing of carbon dioxide.

PubMed

Oloman, Colin; Li, Hui

2008-01-01

With respect to the negative role of carbon dioxide on our climate, it is clear that the time is ripe for the development of processes that convert CO(2) into useful products. The electroreduction of CO(2) is a prime candidate here, as the reaction at near-ambient conditions can yield organics such as formic acid, methanol, and methane. Recent laboratory work on the 100 A scale has shown that reduction of CO(2) to formate (HCO(2)(-)) may be carried out in a trickle-bed continuous electrochemical reactor under industrially viable conditions. Presuming the problems of cathode stability and formate crossover can be overcome, this type of reactor is proposed as the basis for a commercial operation. The viability of corresponding processes for electrosynthesis of formate salts and/or formic acid from CO(2) is examined here through conceptual flowsheets for two process options, each converting CO(2) at the rate of 100 tonnes per day.

Heat transfer in laminar flow along circular rods in infinite square arrays

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

Kim, J.H.; Li, W.H.

1988-02-01

The need to understand heat transfer characteristics over rods or tube bundles often arises in the design of compact heat exchangers and safety analysis of nuclear reactors. In particular, the fuel bundles of typical light water nuclear reactors are composed of a large number of circular rods arranged in square array pattern. The purpose of the present study is to analyze heat transfer characteristics of flow in such a multirod geometric configuration. The analysis given here will follow as closely as possible the method of Sparrow et al. who analyzed a similar problem for circular cylinders arranged in an equilateralmore » triangular array. The following major assumptions are made in the present analysis: (1) Flow is fully developed laminar flow paralleled to the axis of rods. (2) The axial profile of the surface heat flux to the fluid is uniform.(3) Thermodynamic properties are assumed constant.« less

Resonance treatment using pin-based pointwise energy slowing-down method

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

Choi, Sooyoung, E-mail: csy0321@unist.ac.kr; Lee, Changho, E-mail: clee@anl.gov; Lee, Deokjung, E-mail: deokjung@unist.ac.kr

A new resonance self-shielding method using a pointwise energy solution has been developed to overcome the drawbacks of the equivalence theory. The equivalence theory uses a crude resonance scattering source approximation, and assumes a spatially constant scattering source distribution inside a fuel pellet. These two assumptions cause a significant error, in that they overestimate the multi-group effective cross sections, especially for {sup 238}U. The new resonance self-shielding method solves pointwise energy slowing-down equations with a sub-divided fuel rod. The method adopts a shadowing effect correction factor and fictitious moderator material to model a realistic pointwise energy solution. The slowing-down solutionmore » is used to generate the multi-group cross section. With various light water reactor problems, it was demonstrated that the new resonance self-shielding method significantly improved accuracy in the reactor parameter calculation with no compromise in computation time, compared to the equivalence theory.« less

Fast particles in a steady-state compact FNS and compact ST reactor

NASA Astrophysics Data System (ADS)

Gryaznevich, M. P.; Nicolai, A.; Buxton, P.

2014-10-01

This paper presents results of studies of fast particles (ions and alpha particles) in a steady-state compact fusion neutron source (CFNS) and a compact spherical tokamak (ST) reactor with Monte-Carlo and Fokker-Planck codes. Full-orbit simulations of fast particle physics indicate that a compact high field ST can be optimized for energy production by a reduction of the necessary (for the alpha containment) plasma current compared with predictions made using simple analytic expressions, or using guiding centre approximation in a numerical code. Alpha particle losses may result in significant heating and erosion of the first wall, so such losses for an ST pilot plant have been calculated and total and peak wall loads dependence on the plasma current has been studied. The problem of dilution has been investigated and results for compact and big size devices are compared.

Apparatus, components and operating methods for circulating fluidized bed transport gasifiers and reactors

DOEpatents

Vimalchand, Pannalal; Liu, Guohai; Peng, Wan Wang

2015-02-24

The improvements proposed in this invention provide a reliable apparatus and method to gasify low rank coals in a class of pressurized circulating fluidized bed reactors termed "transport gasifier." The embodiments overcome a number of operability and reliability problems with existing gasifiers. The systems and methods address issues related to distribution of gasification agent without the use of internals, management of heat release to avoid any agglomeration and clinker formation, specific design of bends to withstand the highly erosive environment due to high solid particles circulation rates, design of a standpipe cyclone to withstand high temperature gasification environment, compact design of seal-leg that can handle high mass solids flux, design of nozzles that eliminate plugging, uniform aeration of large diameter Standpipe, oxidant injection at the cyclone exits to effectively modulate gasifier exit temperature and reduction in overall height of the gasifier with a modified non-mechanical valve.

Innovative test method for the estimation of the foaming tendency of substrates for biogas plants.

PubMed

Moeller, Lucie; Eismann, Frank; Wißmann, Daniel; Nägele, Hans-Joachim; Zielonka, Simon; Müller, Roland A; Zehnsdorf, Andreas

2015-07-01

Excessive foaming in anaerobic digestion occurs at many biogas plants and can cause problems including plugged gas pipes. Unfortunately, the majority of biogas plant operators are unable to identify the causes of foaming in their biogas reactor. The occurrence of foaming is often related to the chemical composition of substrates fed to the reactor. The consistency of the digestate itself is also a crucial part of the foam formation process. Thus, no specific recommendations concerning substrates can be given in order to prevent foam formation in biogas plants. The safest way to avoid foaming is to test the foaming tendency of substrates on-site. A possible solution is offered by an innovative foaming test. With the help of this tool, biogas plant operators can evaluate the foaming disposition of new substrates prior to use in order to adjust the composition of substrate mixes. Copyright © 2015 Elsevier Ltd. All rights reserved.

Controlled multistep synthesis in a three-phase droplet reactor

PubMed Central

Nightingale, Adrian M.; Phillips, Thomas W.; Bannock, James H.; de Mello, John C.

2014-01-01

Channel-fouling is a pervasive problem in continuous flow chemistry, causing poor product control and reactor failure. Droplet chemistry, in which the reaction mixture flows as discrete droplets inside an immiscible carrier liquid, prevents fouling by isolating the reaction from the channel walls. Unfortunately, the difficulty of controllably adding new reagents to an existing droplet stream has largely restricted droplet chemistry to simple reactions in which all reagents are supplied at the time of droplet formation. Here we describe an effective method for repeatedly adding controlled quantities of reagents to droplets. The reagents are injected into a multiphase fluid stream, comprising the carrier liquid, droplets of the reaction mixture and an inert gas that maintains a uniform droplet spacing and suppresses new droplet formation. The method, which is suited to many multistep reactions, is applied to a five-stage quantum dot synthesis wherein particle growth is sustained by repeatedly adding fresh feedstock. PMID:24797034

Transmutation of Radioactive Nuclear Waste — Present Status and Requirement for the Problem-Oriented Nuclear Database: Approach to Scheduling the Experiments (Reactor, Target, Blanket)

NASA Astrophysics Data System (ADS)

Artisyuk, V.; Ignatyuk, A.; Korovin, Yu.; Lopatkin, A.; Matveenko, I.; Stankovskiy, A.; Titarenko, Yu.

2005-05-01

Transmutation of nuclear wastes (Minor Actinides and Long-Lived Fission Products) remains an important option to reduce the burden of high-level waste on final waste disposal in deep geological structures. Accelerator-Driven Systems (ADS) are considered as possible candidates to perform transmutation due to their subcritical operation mode that eliminates some of the serious safety penalties unavoidable in critical reactors. Specific requirements to nuclear data necessary for ADS transmutation analysis is the main subject of the ISTC Project ♯2578 which started in 2004 to identify the areas of research priorities in the future. The present paper gives a summary of ongoing project stressing the importance of nuclear data for blanket performance (reactivity behavior with associated safety characteristics) and uncertainties that affect characteristics of neutron producing target.

An assessment of coupling algorithms for nuclear reactor core physics simulations

DOE PAGES

Hamilton, Steven; Berrill, Mark; Clarno, Kevin; ...

2016-04-01

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Furthermore, numerical simulations demonstrating the efficiency ofmore » JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.« less

A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377

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

Carelli, M.D.; Franceschini, F.; Lahoda, E.J.

2012-07-01

A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are thatmore » the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)« less