GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE ...

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

GRAPHITE BLOCKS ARE ARRAYED IN "THERMAL COLUMN" ON NORTH SIDE OF REACTOR. INL NEGATIVE NO. 4000. Unknown Photographer, 12/28/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REFRACTORY COATING FOR GRAPHITE MOLDS

DOEpatents

Stoddard, S.D.

1958-06-24

Refractory coating for graphite molds used in the casting of uranium is described. The coating is an alumino-silicate refractory composition which may be used as a mold surface in solid form or as a coating applied to the graphite mold. The composition consists of a mixture of ball clay, kaolin, alumina cement, alumina, water, sodium silicate, and sodium carbonate.

GRAFEC: A New Spanish Program to Investigate Waste Management Options for Radioactive Graphite - 12399

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

Marquez, Eva; Pina, Gabriel; Rodriguez, Marina

Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m{sup 3} each and placed in interimmore » storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like {sup 14}C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of {sup 14}C by thermal methods. This method is based on results obtained at the Research Centre Juelich (FZJ) in the Frame of the EC programs 'Raphael' and 'Carbowaste'. The process developed at FZJ is based on a preferential oxidation of {sup 14}C in comparison to the bulk {sup 12}C. Explanations for this effect are the inhomogeneous distribution and a weaker bounding of {sup 14}C which is not incorporated in the graphite lattice. However these investigations have only been performed with graphite from the high temperature reactor Arbeitsgemeinschaft Versuchsreaktor Juelich AVR which has been operated in a non-oxidising condition or research reactor graphite operated at room temperature. The reactor Vandellos I has been operated with CO{sub 2} as coolant and significant amounts of graphite have been already oxidised. The aim of the project is to validate whether a {sup 14}C decontamination can also been achieved with graphite from Vandellos I. A second possibility under investigation is the encapsulation of the graphite in a long term stable glass matrix. The principal applicability has been already proved by FNAG. Crushed graphite mixed with a suitable glass powder has been pressed at elevated temperature under vacuum. The vacuum is required to avoid gas enclosures in the obtained product. The obtained products, named IGM for 'Impermeable Graphite Matrix', have densities above 99% of theoretical density. The amount of glass has been chosen with respect to the pore volume of the former graphite parts. The method allows the production of encapsulated graphite without increasing the disposal volume. This paper will give a short overview of characterisation results of different irradiated graphite materials obtained at CIEMAT and in the Carbowaste project as well as the proposed methods and the actual status of the program including first results about leaching of non-radioactive IGM samples and hopefully first tendencies concerning the C-14 separation from graphite of Vandellos I by thermal treatment. Both processes, the thermal treatment as well as the IGM, have the potential to solve problems related to the management of irradiated graphite in Spain. However the methods have only been tested with different types of i-graphite and virgin graphite, respectively. Only investigations with real i-graphite from Spain will reveal whether the described methods are applicable to graphite from Vandellos I. However all partners are convinced that one of these new methods or a combination of them will lead to a feasible option to manage i-graphite in Spain on an industrial scale. (authors)« less

REACTOR HAVING NaK-UO$sub 2$ SLURRY HELICALLY POSITIONED IN A GRAPHITE MODERATOR

DOEpatents

Rodin, M.B.; Carter, J.C.

1962-05-15

A reactor utilizing 20% enriched uranium consists of a central graphite island in cylindrical form, with a spiral coil of tubing fitting against the central island. An external graphite moderator is placed around the central island and coil. A slurry of uranium dioxide dispersed in alkali metal passes through the coil to transfer heat externally to the reactor. There are also conventional controls for regulating the nuclear reaction. (AEC)

Neutronic reactor

DOEpatents

Carleton, John T.

1977-01-25

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

MODELING THE ELECTROLYTIC DECHLORINATION OF TRICHLOROETHYLENE IN A GRANULAR GRAPHITE-PACKED REACTOR

EPA Science Inventory

A comprehensive reactor model was developed for the electrolytic dechlorination of trichloroethylene (TCE) at a granular-graphite cathode. The reactor model describes the dynamic processes of TCE dechlorination and adsorption, and the formation and dechlorination of all the major...

EFFECT OF MASSIVE NEUTRON EXPOSURE ON THE DISTORTION OF REACTOR GRAPHITE

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

Helm, J.W.; Davidson, J.M.

1963-05-28

Distortion of reactor-grade graphites was studied at varying neutron exposures ranging up to 14 x 10/sup 21/ neutrons per cm/sup 2/ (nvt)/sup */ at temperatures of irradiation ranging from 425 to 800 deg C. This exposure level corresponds to approximately 100,000 megawatt days per adjacent ton of fuel (Mwd/ At) in a graphite-moderated reactor. A conventionalcoke graphite, CSF, and two needle-coke graphites, NC-7 and NC-8, were studied. At all temperatures of irradiation the contraction rate of the samples cut parallel to the extrusion axis increased with increasing neutron exposure. For parallel samples the needle- coke graphites and the CSF graphitemore » contracted approximately the same amount. In the transverse direction the rate of cortraction at the higher irradiation temperntures appeared to be decreasing. Volume contractions derived from the linear contractions are discussed. (auth)« less

High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

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

Mcwilliams, A. J.

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniquesmore » through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.« less

Examination of Surface Deposits on Oldbury Reactor Core Graphite to Determine the Concentration and Distribution of 14C.

PubMed

Payne, Liam; Heard, Peter J; Scott, Thomas B

2016-01-01

Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C) to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples originating lower in the reactor exhibiting a higher concentration of 14C. Oxidation at 600°C showed that the remaining graphite material consisted of two fractions of 14C, a surface associated fraction and a graphite lattice associated fraction. The results presented correlate well with previous studies on irradiated graphite that suggest there are up to three fractions of 14C; a readily releasable fraction (corresponding to that removed by oxidation at 450°C in this study), a slowly releasable fraction (removed early at 600°C in this study), and an unreleasable fraction (removed later at 600°C in this study).

Examination of Surface Deposits on Oldbury Reactor Core Graphite to Determine the Concentration and Distribution of 14C

PubMed Central

Payne, Liam; Heard, Peter J.; Scott, Thomas B.

2016-01-01

Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C) to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples originating lower in the reactor exhibiting a higher concentration of 14C. Oxidation at 600°C showed that the remaining graphite material consisted of two fractions of 14C, a surface associated fraction and a graphite lattice associated fraction. The results presented correlate well with previous studies on irradiated graphite that suggest there are up to three fractions of 14C; a readily releasable fraction (corresponding to that removed by oxidation at 450°C in this study), a slowly releasable fraction (removed early at 600°C in this study), and an unreleasable fraction (removed later at 600°C in this study). PMID:27706228

Online Oxide Contamination Measurement and Purification Demonstration

NASA Technical Reports Server (NTRS)

Bradley, D. E.; Godfroy, T. J.; Webster, K. L.; Garber, A. E.; Polzin, K. A.; Childers, D. J.

2011-01-01

Liquid metal sodium-potassium (NaK) has advantageous thermodynamic properties indicating its use as a fission reactor coolant for a surface (lunar, martian) power system. A major area of concern for fission reactor cooling systems is system corrosion due to oxygen contaminants at the high operating temperatures experienced. A small-scale, approximately 4-L capacity, simulated fission reactor cooling system employing NaK as a coolant was fabricated and tested with the goal of demonstrating a noninvasive oxygen detection and purification system. In order to generate prototypical conditions in the simulated cooling system, several system components were designed, fabricated, and tested. These major components were a fully-sealed, magnetically-coupled mechanical NaK pump, a graphite element heated reservoir, a plugging indicator system, and a cold trap. All system components were successfully demonstrated at a maximum system flow rate of approximately 150 cc/s at temperatures up to 550 C. Coolant purification was accomplished using a cold trap before and after plugging operations which showed a relative reduction in oxygen content.

Role of nuclear grade graphite in controlling oxidation in modular HTGRs

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

Windes, Willaim; Strydom, G.; Kane, J.

2014-11-01

The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of coremore » environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.« less

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

USGS Publications Warehouse

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

1991-01-01

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

Method of making a sodium sulfur battery

DOEpatents

Elkins, Perry E.

1981-01-01

A method of making a portion of a sodium sulfur battery is disclosed. The battery portion made is a portion of the container which defines the volume for the cathodic reactant materials which are sulfur and sodium polysulfide materials. The container portion is defined by an outer metal casing with a graphite liner contained therein, the graphite liner having a coating on its internal diameter for sealing off the porosity thereof. The steel outer container and graphite pipe are united by a method which insures that at the operating temperature of the battery, relatively low electrical resistance exists between the two materials because they are in intimate contact with one another.

REFLECTOR FOR NEUTRONIC REACTORS

DOEpatents

Fraas, A.P.

1963-08-01

A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

Neutronic reactor thermal shield

DOEpatents

Wende, Charles W. J.

1976-06-15

1. The method of operating a water-cooled neutronic reactor having a graphite moderator which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40-60 volume percent of the mixture, in contact with the graphite moderator.

Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe135 penetration for molten salt breeder reactor

NASA Astrophysics Data System (ADS)

Song, Jinliang; Zhao, Yanling; He, Xiujie; Zhang, Baoliang; Xu, Li; He, Zhoutong; Zhang, DongSheng; Gao, Lina; Xia, Huihao; Zhou, Xingtai; Huai, Ping; Bai, Shuo

2015-01-01

A fixed-bed deposition method was used to prepare rough laminar pyrolytic carbon coating (RLPyC) on graphite for inhibiting liquid fluoride salt and Xe135 penetration during use in molten salt breeder reactor. The RLPyC coating possessed a graphitization degree of 44% and had good contact with graphite substrate. A high-pressure reactor was constructed to evaluate the molten salt infiltration in the isostatic graphite (IG-110, TOYO TANSO CO., LTD.) and RLPyC coated graphite under 1.01, 1.52, 3.04, 5.07 and 10.13 × 105 Pa for 12 h. Mercury injection and molten-salt infiltration experiments indicated the porosity and the salt-infiltration amount of 18.4% and 13.5 wt% under 1.52 × 105 Pa of IG-110, which was much less than 1.2% and 0.06 wt% under 10.13 × 105 Pa of the RLPyC, respectively. A vacuum device was constructed to evaluate the Xe135 penetration in the graphite. The helium diffusion coefficient of RLPyC coated graphite was 2.16 × 10-12 m2/s, much less than 1.21 × 10-6 m2/s of the graphite. Thermal cycle experiment indicated the coatings possessed excellent thermal stability. The coated graphite could effectively inhibit the liquid fluoride salt and Xe135 penetration.

Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

NASA Technical Reports Server (NTRS)

Mireles, Omar R.; Houts, Michael G.

2011-01-01

Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

GUM Analysis for TIMS and SIMS Isotopic Ratios in Graphite

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

Heasler, Patrick G.; Gerlach, David C.; Cliff, John B.

2007-04-01

This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

AGC-2 Graphite Pre-irradiation Data Package

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

David Swank; Joseph Lord; David Rohrbaugh

2010-08-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterizedmore » prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

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

2017-04-01

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

Method of making a sodium sulfur battery

DOEpatents

Elkins, P. E.

1981-09-22

A method of making a portion of a sodium sulfur battery is disclosed. The battery portion made is a portion of the container which defines the volume for the cathodic reactant materials which are sulfur and sodium polysulfide materials. The container portion is defined by an outer metal casing with a graphite liner contained therein, the graphite liner having a coating on its internal diameter for sealing off the porosity thereof. The steel outer container and graphite pipe are united by a method which insures that at the operating temperature of the battery, relatively low electrical resistance exists between the two materials because they are in intimate contact with one another. 3 figs.

The effect of carbon crystal structure on treat reactor physics calculations

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

Swanson, R.W.; Harrison, L.J.

1988-01-01

The Transient Reactor Test Facility (TREAT) at Argonne National Laboratory-West (ANL-W) is fueled with urania in a graphite and carbon mixture. This fuel was fabricated from a mixture of graphite flour, thermax (a thermatomic carbon produced by ''cracking'' natural gas), coal-tar resin and U/sub 3/O/sub 8/. During the fabrication process, the fuel was baked to dissociate the resin, but the high temperature necessary to graphitize the carbon in the thermax and in the resin was avoided. Therefore, the carbon crystal structure is a complex mixture of graphite particles in a nongraphitized elemental carbon matrix. Results of calculations using macroscopic carbonmore » cross sections obtained by mixing bound-kernel graphite cross sections for the graphitized carbon and free-gas carbon cross sections for the remainder of the carbon and calculations using only bound-kernel graphite cross sections are compared to experimental data. It is shown that the use of the hybridized cross sections which reflect the allotropic mixture of the carbon in the TREAT fuel results in a significant improvement in the accuracy of calculated neutronics parameters for the TREAT reactor. 6 refs., 2 figs., 3 tabs.« less

GUM Analysis for SIMS Isotopic Ratios in BEP0 Graphite Qualification Samples, Round 2

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

Gerlach, David C.; Heasler, Patrick G.; Reid, Bruce D.

2009-01-01

This report describes GUM calculations for TIMS and SIMS isotopic ratio measurements of reactor graphite samples. These isotopic ratios are used to estimate reactor burn-up, and currently consist of various ratios of U, Pu, and Boron impurities in the graphite samples. The GUM calculation is a propagation of error methodology that assigns uncertainties (in the form of standard error and confidence bound) to the final estimates.

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.

EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE

EPA Science Inventory

Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE.

EPA Science Inventory

Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...

THE FUEL ELEMENT GRAPHITE. Project DRAGON.

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

Graham, L.W.; Price, M.S.T.

1963-01-15

The main requirements of a fuel element graphite for reactors based on the Dragon concept are low transmission coefficient for fission products, dimensional stability under service conditions, high strength, high thermal conductivity, high purity, and high resistance to oxidation. Since conclusions reached in early 1960, a considerable amount of information has accumulated concerning the likely behaviour of graphites in high temperature reactor systems, particularly data on dimensional stability under irradiation. The influence of this new knowledge on the development of fuel element graphite with the Dragon Project is discussed in detail in the final section of this paper.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE PAGES

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

2017-02-04

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

Effect of reactor radiation on the thermal conductivity of TREAT fuel

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

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

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

ELECTROCHEMICAL DEGRADATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODES: IDENTIFICATION AND QUALIFICATION OF DECHLORINATION PRODUCTS

EPA Science Inventory

TCE was successfully dechlorinated in aqueous solution using granular graphite as the cathode in a mixed electrochemical reactor. In experiments with an initial TCE concentration of less than 100 mg/l, TCE was reduced approximately by 75% in the reactor under an applied cell volt...

New insights into canted spiro carbon interstitial in graphite

NASA Astrophysics Data System (ADS)

EL-Barbary, A. A.

2017-12-01

The self-interstitial carbon is the key to radiation damage in graphite moderator nuclear reactor, so an understanding of its behavior is essential for plant safety and maximized reactor lifetime. The density functional theory is applied on four different graphite unit cells, starting from of 64 carbon atoms up to 256 carbon atoms, using AIMPRO code to obtain the energetic, athermal and mechanical properties of carbon interstitial in graphite. This study presents first principles calculations of the energy of formation that prove its high barrier to athermal diffusion (1.1 eV) and the consequent large critical shear stress (39 eV-50 eV) necessary to shear graphite planes in its presence. Also, for the first time, the gamma surface of graphite in two dimensions is calculated and found to yield the critical shear stress for perfect graphite. Finally, in contrast to the extensive literature describing the interstitial of carbon in graphite as spiro interstitial, in this work the ground state of interstitial carbon is found to be canted spiro interstitial.

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

Mohanty, Subhasish; Majumdar, Saurindranath

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

Nondestructive evaluation of nuclear-grade graphite

NASA Astrophysics Data System (ADS)

Kunerth, D. C.; McJunkin, T. R.

2012-05-01

The material of choice for the core of the high-temperature gas-cooled reactors being developed by the U.S. Department of Energy's Next Generation Nuclear Plant Program is graphite. Graphite is a composite material whose properties are highly dependent on the base material and manufacturing methods. In addition to the material variations intrinsic to the manufacturing process, graphite will also undergo changes in material properties resulting from radiation damage and possible oxidation within the reactor. Idaho National Laboratory is presently evaluating the viability of conventional nondestructive evaluation techniques to characterize the material variations inherent to manufacturing and in-service degradation. Approaches of interest include x-ray radiography, eddy currents, and ultrasonics.

DR Reactor VSR channel damage

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

Kempf, F.J.; Rawlins, J.K.

1961-10-30

On July 11, 1961 the Ball 3X System at DR Reactor was inadventently tripped. All vertical safety rods dropped and all channels were filled with balls. This report has the twofold purpose of documenting borescope observations of ten vertical rod channels at DR Reactor and recording the estimated extent of graphite damage resulting from the above incident. Channel damage data are presented on appended drawings. With suitable notations, the tracings of these drawings may be revised to reflect any future graphite damage. All vertical rod channels at DR Reactor were visually examined with a closed circuit television system during ballmore » removal efforts. Typical photographs of trapped balls and ledges, as viewed on the television monitor, are shown. Photographs of typical graphite damage, obtained through the borescope are also included in this report. 3 refs., 8 figs., 1 tab.« less

Synthesis and Characterization of Highly Intercalated Graphite Bisulfate

NASA Astrophysics Data System (ADS)

Salvatore, Marcella; Carotenuto, Gianfranco; De Nicola, Sergio; Camerlingo, Carlo; Ambrogi, Veronica; Carfagna, Cosimo

2017-03-01

Different chemical formulations for the synthesis of highly intercalated graphite bisulfate have been tested. In particular, nitric acid, potassium nitrate, potassium dichromate, potassium permanganate, sodium periodate, sodium chlorate, and hydrogen peroxide have been used in this synthesis scheme as the auxiliary reagent (oxidizing agent). In order to evaluate the presence of delamination, and pre-expansion phenomena, and the achieved intercalation degree in the prepared samples, the obtained graphite intercalation compounds have been characterized by scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), X-ray powder diffraction (XRD), infrared spectroscopy (FT-IR), micro-Raman spectroscopy ( μ-RS), and thermal analysis (TGA). Delamination and pre-expansion phenomena were observed only for nitric acid, sodium chlorate, and hydrogen peroxide, while the presence of strong oxidizers (KMnO4, K2Cr2O7) led to stable graphite intercalation compounds. The largest content of intercalated bisulfate is achieved in the intercalated compounds obtained from NaIO4 and NaClO3.

Synthesis and Characterization of Highly Intercalated Graphite Bisulfate.

PubMed

Salvatore, Marcella; Carotenuto, Gianfranco; De Nicola, Sergio; Camerlingo, Carlo; Ambrogi, Veronica; Carfagna, Cosimo

2017-12-01

Different chemical formulations for the synthesis of highly intercalated graphite bisulfate have been tested. In particular, nitric acid, potassium nitrate, potassium dichromate, potassium permanganate, sodium periodate, sodium chlorate, and hydrogen peroxide have been used in this synthesis scheme as the auxiliary reagent (oxidizing agent). In order to evaluate the presence of delamination, and pre-expansion phenomena, and the achieved intercalation degree in the prepared samples, the obtained graphite intercalation compounds have been characterized by scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), X-ray powder diffraction (XRD), infrared spectroscopy (FT-IR), micro-Raman spectroscopy (μ-RS), and thermal analysis (TGA). Delamination and pre-expansion phenomena were observed only for nitric acid, sodium chlorate, and hydrogen peroxide, while the presence of strong oxidizers (KMnO 4 , K 2 Cr 2 O 7 ) led to stable graphite intercalation compounds. The largest content of intercalated bisulfate is achieved in the intercalated compounds obtained from NaIO 4 and NaClO 3 .

Coating method for graphite

DOEpatents

Banker, John G.; Holcombe, Jr., Cressie E.

1977-01-01

A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided comprising coating the graphite surface with a suspension of Y.sub.2 O.sub.3 particles in water containing about 1.5 to 4% by weight sodium carboxymethylcellulose.

Coating method for graphite

DOEpatents

Banker, J.G.; Holcombe, C.E. Jr.

1975-11-06

A method of limiting carbon contamination from graphite ware used in induction melting of uranium alloys is provided. The graphite surface is coated with a suspension of Y/sub 2/O/sub 3/ particles in water containing about 1.5 to 4 percent by weight sodium carboxymethylcellulose.

Removal of metals from aqueous solution and sea water by functionalized graphite nanoplatelets based electrodes.

PubMed

Mishra, Ashish Kumar; Ramaprabhu, S

2011-01-15

In the present wok, we have demonstrated the simultaneous removal of sodium and arsenic (pentavalent and trivalent) from aqueous solution using functionalized graphite nanoplatelets (f-GNP) based electrodes. In addition, these electrodes based water filter was used for multiple metals removal from sea water. Graphite nanoplatelets (GNP) were prepared by acid intercalation and thermal exfoliation. Functionalization of GNP was done by further acid treatment. Material was characterized by different characterization techniques. Performance of supercapacitor based water filter was analyzed for the removal of high concentration of arsenic (trivalent and pentavalent) and sodium as well as for desalination of sea water, using cyclic voltametry (CV) and inductive coupled plasma-optical emission spectroscopy (ICP-OES) techniques. Adsorption isotherms and kinetic characteristics were studied for the simultaneous removal of sodium and arsenic (both trivalent and pentavalent). Maximum adsorption capacities of 27, 29 and 32 mg/g for arsenate, arsenite and sodium were achieved in addition to good removal efficiency for sodium, magnesium, calcium and potassium from sea water. Copyright © 2010 Elsevier B.V. All rights reserved.

ICP-MS analysis of fission product diffusion in graphite for High-Temperature Gas-Cooled Reactors

NASA Astrophysics Data System (ADS)

Carter, Lukas M.

Release of radioactive fission products from nuclear fuel during normal reactor operation or in accident scenarios is a fundamental safety concern. Of paramount importance are the understanding and elucidation of mechanisms of chemical interaction, nuclear interaction, and transport phenomena involving fission products. Worldwide efforts to reduce fossil fuel dependence coupled with an increasing overall energy demand have generated renewed enthusiasm toward nuclear power technologies, and as such, these mechanisms continue to be the subjects of vigorous research. High-Temperature Gas-Cooled Reactors (HTGRs or VHTRs) remain one of the most promising candidates for the next generation of nuclear power reactors. An extant knowledge gap specific to HTGR technology derives from an incomplete understanding of fission product transport in major core materials under HTGR operational conditions. Our specific interest in the current work is diffusion in reactor graphite. Development of methods for analysis of diffusion of multiple fission products is key to providing accurate models for fission product release from HTGR core components and the reactor as a whole. In the present work, a specialized diffusion cell has been developed and constructed to facilitate real-time diffusion measurements via ICP-MS. The cell utilizes a helium gas-jet system which transports diffusing fission products to the mass spectrometer using carbon nanoparticles. The setup was designed to replicate conditions present in a functioning HTGR, and can be configured for real-time release or permeation measurements of single or multiple fission products from graphite or other core materials. In the present work, we have analyzed release rates of cesium in graphite grades IG-110, NBG-18, and a commercial grade of graphite, as well as release of iodine in IG-110. Additionally we have investigated infusion of graphite samples with Cs, I, Sr, Ag, and other surrogate fission products for use in release or profile measurements of diffusion coefficients.

Process for making silicon from halosilanes and halosilicons

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1988-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

Process for making silicon

NASA Technical Reports Server (NTRS)

Levin, Harry (Inventor)

1987-01-01

A reactor apparatus (10) adapted for continuously producing molten, solar grade purity elemental silicon by thermal reaction of a suitable precursor gas, such as silane (SiH.sub.4), is disclosed. The reactor apparatus (10) includes an elongated reactor body (32) having graphite or carbon walls which are heated to a temperature exceeding the melting temperature of silicon. The precursor gas enters the reactor body (32) through an efficiently cooled inlet tube assembly (22) and a relatively thin carbon or graphite septum (44). The septum (44), being in contact on one side with the cooled inlet (22) and the heated interior of the reactor (32) on the other side, provides a sharp temperature gradient for the precursor gas entering the reactor (32) and renders the operation of the inlet tube assembly (22) substantially free of clogging. The precursor gas flows in the reactor (32) in a substantially smooth, substantially axial manner. Liquid silicon formed in the initial stages of the thermal reaction reacts with the graphite or carbon walls to provide a silicon carbide coating on the walls. The silicon carbide coated reactor is highly adapted for prolonged use for production of highly pure solar grade silicon. Liquid silicon (20) produced in the reactor apparatus (10) may be used directly in a Czochralski or other crystal shaping equipment.

Analyzing the impact of reactive transport on the repository performance of TRISO fuel

NASA Astrophysics Data System (ADS)

Schmidt, Gregory

One of the largest determiners of the amount of electricity generated by current nuclear reactors is the efficiency of the thermodynamic cycle used for power generation. Current light water reactors (LWR) have an efficiency of 35% or less for the conversion of heat energy generated by the reactor to electrical energy. If this efficiency could be improved, more power could be generated from equivalent volumes of nuclear fuel. One method of improving this efficiency is to use a coolant flow that operates at a much higher temperature for electricity production. A reactor design that is currently proposed to take advantage of this efficiency is a graphite-moderated, helium-cooled reactor known as a High Temperature Gas Reactor (HTGR). There are significant differences between current LWR's and the proposed HTGR's but most especially in the composition of the nuclear fuel. For LWR's, the fuel elements consist of pellets of uranium dioxide or plutonium dioxide that are placed in long tubes made of zirconium metal alloys. For HTGR's, the fuel, known as TRISO (TRIstructural-ISOtropic) fuel, consists of an inner sphere of fissile material, a layer of dense pyrolytic carbon (PyC), a ceramic layer of silicon carbide (SiC) and a final dense outer layer of PyC. These TRISO particles are then compacted with graphite into fuel rods that are then placed in channels in graphite blocks. The blocks are then arranged in an annular fashion to form a reactor core. However, this new fuel form has unanswered questions on the environmental post-burn-up behavior. The key question for current once-through fuel operations is how these large irradiated graphite blocks with spent fuel inside will behave in a repository environment. Data in the literature to answer this question is lacking, but nevertheless this is an important question that must be answered before wide-spread adoption of HTGR's could be considered. This research has focused on answering the question of how the large quantity of graphite surrounding the spent HTGR fuel will impact the release of aqueous uranium from the TRISO fuel. In order to answer this question, the sorption and partitioning behavior of uranium to graphite under a variety of conditions was investigated. Key systematic variables that were analyzed include solution pH, dissolved carbonate concentration, uranium metal concentration and ionic strength. The kinetics and desorption characteristics of uranium/graphite partitioning were studied as well. The graphite used in these experiments was also characterized by a variety of techniques and conclusions are drawn about the relevant surface chemistry of graphite. This data was then used to generate a model for the reactive transport of uranium in a graphite matrix. This model was implemented with the software code CXTFIT and validated through the use of column studies mirroring the predicted system.

Fabrication of TREAT Fuel with Increased Graphite Loading

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

Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.

2014-02-05

As part of the feasibility study exploring the replacement of the HEU fuel core of the TREAT reactor at Idaho National Laboratory with LEU fuel, this study demonstrates that it is possible to increase the graphite content of extruded fuel by reformulation. The extrusion process was use to fabricate the “upgrade” core1 for the TREAT reactor. The graphite content achieved is determined by calculation and has not been measured by any analytical method. In conjunction, a technique, Raman Spectroscopy, has been investigated for measuring the graphite content. This method shows some promise in differentiating between carbon and graphite; however, standardsmore » that would allow the technique to be calibrated to quantify the graphite concentration have yet to be fabricated. Continued research into Raman Spectroscopy is on going. As part of this study, cracking of graphite extrusions due to volatile evolution during heat treatment has been largely eliminated. Continued research to optimize this extrusion method is required.« less

ICP-MS measurement of iodine diffusion in IG-110 graphite for HTGR/VHTR

NASA Astrophysics Data System (ADS)

Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2016-05-01

Graphite functions as a structural material and as a barrier to fission product release in HTGR/VHTR designs, and elucidation of transport parameters for fission products in reactor-grade graphite is thus required for reactor source terms calculations. We measured iodine diffusion in spheres of IG-110 graphite using a release method based on Fickain diffusion kinetics. Two sources of iodine were loaded into the graphite spheres; molecular iodine (I2) and cesium iodide (CsI). Measurements of the diffusion coefficient were made over a temperature range of 873-1293 K. We have obtained the following Arrhenius expressions for iodine diffusion:DI , CsI infused =(6 ×10-12 2/s) exp(30,000 J/mol RT) And,DI , I2 infused =(4 ×10-10 m2/s) exp(-11,000 J/mol RT ) The results indicate that iodine diffusion in IG-110 graphite is not well-described by Fickan diffusion kinetics. To our knowledge, these are the first measurements of iodine diffusion in IG-110 graphite.

Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

NASA Astrophysics Data System (ADS)

Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

2017-12-01

The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

Pillared graphite anodes for reversible sodiation.

PubMed

Zhang, Hanyang; Li, Zhifei; Xu, Wei; Chen, Yicong; Ji, Xiulei; Lerner, Michael M

2018-08-10

There has been a major effort recently to develop new rechargeable sodium-ion electrodes. In lithium ion batteries, LiC 6 forms from graphite and desolvated Li cations during the first charge. With sodium ions, graphite only shows a significant capacity when Na + intercalates as a solvated complex, resulting in ternary graphite intercalation compounds (GICs). Although this chemistry has been shown to be highly reversible and to support high rates in small test cells, these GICs can require >250% volume expansion and contraction during cycling. Here we demonstrate the first example of GICs that reversibly sodiate/desodiate without any significant volume change. These pillared GICs are obtained by electrochemical reduction of graphite in an ether/amine co-solvent electrolyte. The initial gallery expansion, 0.36 nm, is less than half of that in diglyme-based systems, and shows a similar capacity. Thermal analyses suggest the pillaring phenomenon arises from stronger co-intercalate interactions in the GIC galleries.

FUEL ASSAY REACTOR

DOEpatents

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

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

DTIC Science & Technology

1989-06-01

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

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

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

Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.

A thorough understanding of oxidation is important when considering the health and integrity of graphite components in graphite reactors. For the next generation of graphite reactors, HTGRs specifically, an unlikely air ingress has been deemed significant enough to have made its way into the licensing applications of many international licensing bodies. While a substantial body of literature exists on nuclear graphite oxidation in the presence of molecular oxygen and significant efforts have been made to characterize oxidation kinetics of various grades, the value of existing information is somewhat limited. Often, multiple competing processes, including reaction kinetics, mass transfer, and microstructuralmore » evolution, are lumped together into a single rate expression that limits the ability to translate this information to different conditions. This article reviews the reaction of graphite with molecular oxygen in terms of the reaction kinetics, gas transport, and microstructural evolution of graphite. It also presents the foundations of a model for the graphite-molecular oxygen reaction system that is kinetically independent of graphite grade, and is capable of describing both the bulk and local oxidation rates under a wide range of conditions applicable to air-ingress.« less

Understanding the reaction of nuclear graphite with molecular oxygen: Kinetics, transport, and structural evolution

DOE PAGES

Kane, Joshua J.; Contescu, Cristian I.; Smith, Rebecca E.; ...

2017-06-08

A thorough understanding of oxidation is important when considering the health and integrity of graphite components in graphite reactors. For the next generation of graphite reactors, HTGRs specifically, an unlikely air ingress has been deemed significant enough to have made its way into the licensing applications of many international licensing bodies. While a substantial body of literature exists on nuclear graphite oxidation in the presence of molecular oxygen and significant efforts have been made to characterize oxidation kinetics of various grades, the value of existing information is somewhat limited. Often, multiple competing processes, including reaction kinetics, mass transfer, and microstructuralmore » evolution, are lumped together into a single rate expression that limits the ability to translate this information to different conditions. This article reviews the reaction of graphite with molecular oxygen in terms of the reaction kinetics, gas transport, and microstructural evolution of graphite. It also presents the foundations of a model for the graphite-molecular oxygen reaction system that is kinetically independent of graphite grade, and is capable of describing both the bulk and local oxidation rates under a wide range of conditions applicable to air-ingress.« less

PALLADIUM-FACILITATED ELECTROLYTIC DECHLORINATION OF 2-CHLOROBIPHENYL USING A GRANULAR-GRAPHITE ELECTRODE.

EPA Science Inventory

Palladium-assisted electrocatalytic dechlorination of 2-chlorobiphenyl (2-Cl BP) in aqueous solutions was conducted in a membrane-separated electrochemical reactor with granular-graphite packed electrodes. The dechlorination took place at a granular-graphite cathode while Pd was ...

R and D program for French sodium fast reactor: On the description and detection of sodium boiling phenomena during sub-assembly blockages

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

Vanderhaegen, M.; Laboratory of Waves and Acoustic, Institut Langevin, ESPCI ParisTech, 10 rue Vauquelin, 75005 Paris; Paumel, K.

2011-07-01

In support of the French ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) reactor program, which aims to demonstrate the industrial applicability of sodium fast reactors with an increased level of safety demonstration and availability compared to the past French sodium fast reactors, emphasis is placed on reactor instrumentation. It is in this framework that CEA studies continuous core monitoring to detect as early as possible the onset of sodium boiling. Such a detection system is of particular interest due to the rapid progress and the consequences of a Total Instantaneous Blockage (TIB) at a subassembly inlet, where sodium boilingmore » intervenes in an early phase. In this paper, the authors describe all the particularities which intervene during the different boiling stages and explore possibilities for their detection. (authors)« less

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

DOE PAGES

Contescu, Cristian I.; Mee, Robert W.; Lee, Yoonjo; ...

2017-11-03

Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation withmore » temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.« less

Beyond the classical kinetic model for chronic graphite oxidation by moisture in high temperature gas-cooled reactors

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

Contescu, Cristian I.; Mee, Robert W.; Lee, Yoonjo

Four grades of nuclear graphite with various microstructures were subjected to accelerated oxidation tests in helium with traces of moisture and hydrogen in order to evaluate the effects of chronic oxidation on graphite components in high temperature gas cooled reactors. Kinetic analysis showed that the Langmuir-Hinshelwood (LH) model cannot consistently reproduce all results. In particular, at high temperatures and water partial pressures oxidation was always faster than the LH model predicts, with stronger deviations for superfine grain graphite than for medium grain grades. It was also found empirically that the apparent reaction order for water has a sigmoid-type variation withmore » temperature which follows the integral Boltzmann distribution function. This suggests that the apparent activation with temperature of graphite reactive sites that causes deviations from the LH model is rooted in specific structural and electronic properties of surface sites on graphite. A semi-global kinetic model was proposed, whereby the classical LH model was modified with a temperature-dependent reaction order for water. The new Boltzmann-enhanced model (BLH) was shown to consistently predict experimental oxidation rates over large ranges of temperature (800-1100 oC) and partial pressures of water (3-1200 Pa) and hydrogen (0-300 Pa), not only for the four grades of graphite but also for the historic grade H-451. The BLH model offers as more reliable input for modeling the chemical environment effects during the life-time operation of new grades of graphite in advanced nuclear reactors operating at high and very high temperatures.« less

NUCLEAR REACTORS

DOEpatents

Long, E.; Ashley, J.W.

1958-12-16

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

Control of reactor coolant flow path during reactor decay heat removal

DOEpatents

Hunsbedt, Anstein N.

1988-01-01

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

Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

NASA Astrophysics Data System (ADS)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2017-10-01

Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic impact is lower and can therefore be counteracted by temperature, a better reordering of the structure should be achieved. Concerning 14C, except when located close to open pores where it can be removed through radiolytic corrosion, it tends to stabilize in the graphite matrix into sp2 or sp3 structures with variable proportions depending on the irradiation conditions.

A COOLED NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.; Creutz, E.C.

1960-03-15

A nuclear reactor comprising a pair of graphite blocks separated by an air gap is described. Each of the blocks contains a plurality of channels extending from the gap through the block with a plurality of fuel elements being located in the channels. Means are provided for introducing air into the gap between the graphite blocks and for exhausting the air from the ends of the channels opposite the gap.

Neutronic reactor

DOEpatents

Lewis, Warren R.

1978-05-30

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

PRELIMINARY RESULTS OF THE AGC-4 IRRADIATION IN THE ADVANCED TEST REACTOR AND DESIGN OF AGC-5 (HTR16-18469)

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

Davenport, Michael; Petti, D. A.

The United States Department of Energy’s Advanced Reactor Technologies (ART) Program will irradiate up to six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments are being irradiated over an approximate eight year period to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Very High Temperature Gasmore » Reactor (VHTR), as well as other future gas reactors. The experiments each consist of a single capsule that contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens are not be subjected to a compressive load during irradiation. The six stacks have differing compressive loads applied to the top half of diametrically opposite pairs of specimen stacks. A seventh specimen stack in the center of the capsule does not have a compressive load. The specimens are being irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There are also samples taken of the sweep gas effluent to measure any oxidation or off-gassing of the specimens that may occur during initial start-up of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. The third experiment, AGC-3, started its irradiation in late November 2012 and completed in the April of 2014. AGC-4 is currently being irradiated in the ATR. This paper will briefly discuss the preliminary irradiation results of the AGC-4 experiment, as well as the design of AGC-5.« 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

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

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

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

2007-07-01

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

Systems and methods for dismantling a nuclear reactor

DOEpatents

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

2014-10-28

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

Carbon-14 bioassay for decommissioning of Hanford reactors.

PubMed

Carbaugh, Eugene H; Watson, David J

2012-05-01

The production reactors at the U.S. Department of Energy Hanford Site used large graphite piles as the moderator. As part of long-term decommissioning plans, the potential need for ¹⁴C radiobioassay of workers was identified. Technical issues associated with ¹⁴C bioassay and worker monitoring were investigated, including anticipated graphite characterization, potential intake scenarios, and the bioassay capabilities that may be required to support the decommissioning of the graphite piles. A combination of urine and feces sampling would likely be required for the absorption type S ¹⁴C anticipated to be encountered. However, the concentrations in the graphite piles appear to be sufficiently low that dosimetrically significant intakes of ¹⁴C are not credible, thus rendering moot the need for such bioassay.

Carbon-14 Bioassay for Decommissioning of Hanford Reactors

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

Carbaugh, Eugene H.; Watson, David J.

2012-05-01

The old production reactors at the US Department of Energy Hanford Site used large graphite piles as the moderator. As part of long-term decommissioning plans, the potential need for 14C radiobioassay of workers was identified. Technical issues associated with 14C bioassay and worker monitoring were investigated, including anticipated graphite characterization, potential intake scenarios, and the bioassay capabilities that may be required to support the decommissioning of the graphite piles. A combination of urine and feces sampling would likely be required for the absorption type S 14C anticipated to be encountered. However the concentrations in the graphite piles appear to bemore » sufficiently low that dosimetrically significant intakes of 14C are not credible, thus rendering moot the need for such bioassay.« less

Kinetics of Chronic Oxidation of NBG-17 Nuclear Graphite by Water Vapor

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

Contescu, Cristian I; Burchell, Timothy D; Mee, Robert

2015-05-01

This report presents the results of kinetic measurements during accelerated oxidation tests of NBG-17 nuclear graphite by low concentration of water vapor and hydrogen in ultra-high purity helium. The objective is to determine the parameters in the Langmuir-Hinshelwood (L-H) equation describing the oxidation kinetics of nuclear graphite in the helium coolant of high temperature gas-cooled reactors (HTGR). Although the helium coolant chemistry is strictly controlled during normal operating conditions, trace amounts of moisture (predictably < 0.2 ppm) cannot be avoided. Prolonged exposure of graphite components to water vapor at high temperature will cause very slow (chronic) oxidation over the lifetimemore » of graphite components. This behavior must be understood and predicted for the design and safe operation of gas-cooled nuclear reactors. The results reported here show that, in general, oxidation by water of graphite NBG-17 obeys the L-H mechanism, previously documented for other graphite grades. However, the characteristic kinetic parameters that best describe oxidation rates measured for graphite NBG-17 are different than those reported previously for grades H-451 (General Atomics, 1978) and PCEA (ORNL, 2013). In some specific conditions, certain deviations from the generally accepted L-H model were observed for graphite NBG-17. This graphite is manufactured in Germany by SGL Carbon Group and is a possible candidate for the fuel elements and reflector blocks of HTGR.« less

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

DOE PAGES

Lai, Shigang; Shi, Li; Fok, Alex; ...

2017-01-01

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

A New Method to Measure Crack Extension in Nuclear Graphite Based on Digital Image Correlation

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

Lai, Shigang; Shi, Li; Fok, Alex

Graphite components, used as moderators, reflectors, and core-support structures in a High-Temperature Gas-Cooled Reactor, play an important role in the safety of the reactor. Specifically, they provide channels for the fuel elements, control rods, and coolant flow. Fracture is the main failure mode for graphite, and breaching of the above channels by crack extension will seriously threaten the safety of a reactor. In this paper, a new method based on digital image correlation (DIC) is introduced for measuring crack extension in brittle materials. Cross-correlation of the displacements measured by DIC with a step function was employed to identify the advancingmore » crack tip in a graphite beam specimen under three-point bending. The load-crack extension curve, which is required for analyzing the R-curve and tension softening behaviors, was obtained for this material. Furthermore, a sensitivity analysis of the threshold value employed for the cross-correlation parameter in the crack identification process was conducted. Finally, the results were verified using the finite element method.« less

A sodium-ion battery exploiting layered oxide cathode, graphite anode and glyme-based electrolyte

NASA Astrophysics Data System (ADS)

Hasa, Ivana; Dou, Xinwei; Buchholz, Daniel; Shao-Horn, Yang; Hassoun, Jusef; Passerini, Stefano; Scrosati, Bruno

2016-04-01

Room-temperature rechargeable sodium-ion batteries (SIBs), in view of the large availability and low cost of sodium raw materials, represent an important class of electrochemical systems suitable for application in large-scale energy storage. In this work, we report a novel, high power SIB formed by coupling the layered P2-Na0.7CoO2 cathode with the graphite anode in an optimized ether-based electrolyte. The study firstly addresses the electrochemical optimization of the two electrode materials and then the realization and characterization of the novel SIB based on their combination. The cell represents an original sodium rocking chair battery obtained combining the intercalation/de-intercalation processes of sodium within the cathode and anode layers. We show herein that this battery, favored by suitable electrode/electrolyte combination, offers unique performance in terms of cycle life, efficiency and, especially, power capability.

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

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

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

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

Comparison of irradiation behaviour of HTR graphite grades

NASA Astrophysics Data System (ADS)

Heijna, M. C. R.; de Groot, S.; Vreeling, J. A.

2017-08-01

The INNOGRAPH irradiations were executed in the High Flux Reactor (HFR) in Petten by NRG supported by the European Framework programs HTR-M, RAPHAEL, and ARCHER to generate data on the irradiation behaviour of graphite grades for High Temperature Reactor (HTR) application available at that time. Samples of the graphite grades NBG-10, NBG-17, NBG-18, NBG-20, NBG-25, PCEA, PPEA, PCIB, and IG-110 have been irradiated at 750 °C and 950 °C. The inherent scatter induced by the probabilistic material behaviour of graphite requires uncertainty and scatter induced by test conditions and post-irradiation examination to be minimized. The INNOGRAPH irradiations supplied an adequate number of irradiated samples to enable accurate determination of material properties and their evolution under irradiation. This allows comparison of different graphite grades and a qualitative assessment of their appropriateness for HTR applications, as a basis of selection, design and core component lifetime. The results indicate that coarse grained graphite grades exhibit more favourable behaviour for application in HTRs due to their low dimensional anisotropy and fracture propagation resilience.

Sealing nuclear graphite with pyrolytic carbon

NASA Astrophysics Data System (ADS)

Feng, Shanglei; Xu, Li; Li, Li; Bai, Shuo; Yang, Xinmei; Zhou, Xingtai

2013-10-01

Pyrolytic carbon (PyC) coatings were deposited on IG-110 nuclear graphite by thermal decomposition of methane at ∼1830 °C. The PyC coatings are anisotropic and airtight enough to protect IG-110 nuclear graphite against the permeation of molten fluoride salts and the diffusion of gases. The investigations indicate that the sealing nuclear graphite with PyC coating is a promising method for its application in Molten Salt Reactor (MSR).

NEUTRONIC REACTOR SHIELDING

DOEpatents

Borst, L.B.

1961-07-11

A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

NUCLEAR REACTORS

DOEpatents

Long, E.; Ashby, J.W.

1958-09-16

ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

Fluorine interaction with defects on graphite surface by a first-principles study

NASA Astrophysics Data System (ADS)

Wang, Song; Xuezhi, Ke; Zhang, Wei; Gong, Wenbin; Huai, Ping; Zhang, Wenqing; Zhu, Zhiyuan

2014-02-01

The interaction between fluorine atom and graphite surface has been investigated in the framework of density functional theory. Due to the consideration of molten salt reactor system, only carbon adatoms and vacancies are chemical reactive for fluorine atoms. Fluorine adsorption on carbon adatom will enhance the mobility of carbon adatom. Carbon adatom can also be removed easily from graphite surface in form of CF2 molecule, explaining the formation mechanism of CF2 molecule in previous experiment. For the interaction between fluorine and vacancy, we find that fluorine atoms which adsorb at vacancy can hardly escape. Both pristine surface and vacancy are impossible for fluorine to penetrate due to the high penetration barrier. We believe our result is helpful to understand the compatibility between graphite and fluorine molten salt in molten salt reactor system.

Advanced Na-NiCl2 Battery Using Nickel-Coated Graphite with Core-Shell Microarchitecture.

PubMed

Chang, Hee-Jung; Canfield, Nathan L; Jung, Keeyoung; Sprenkle, Vincent L; Li, Guosheng

2017-04-05

Stationary electric energy storage devices (rechargeable batteries) have gained increasing prominence due to great market needs, such as smoothing the fluctuation of renewable energy resources and supporting the reliability of the electric grid. With regard to raw materials availability, sodium-based batteries are better positioned than lithium batteries due to the abundant resource of sodium in Earth's crust. However, the sodium-nickel chloride (Na-NiCl 2 ) battery, one of the most attractive stationary battery technologies, is hindered from further market penetration by its high material cost (Ni cost) and fast material degradation at its high operating temperature. Here, we demonstrate the design of a core-shell microarchitecture, nickel-coated graphite, with a graphite core to maintain electrochemically active surface area and structural integrity of the electron percolation pathway while using 40% less Ni than conventional Na-NiCl 2 batteries. An initial energy density of 133 Wh/kg (at ∼C/4) and energy efficiency of 94% are achieved at an intermediate temperature of 190 °C.

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

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

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

2016-06-01

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

A study of the relationship between microstructure and oxidation effects in nuclear graphite at very high temperatures

NASA Astrophysics Data System (ADS)

Lo, I.-Hsuan; Tzelepi, Athanasia; Patterson, Eann A.; Yeh, Tsung-Kuang

2018-04-01

Graphite is used in the cores of gas-cooled reactors as both the neutron moderator and a structural material, and traditional and novel graphite materials are being studied worldwide for applications in Generation IV reactors. In this study, the oxidation characteristics of petroleum-based IG-110 and pitch-based IG-430 graphite pellets in helium and air environments at temperatures ranging from 700 to 1600 °C were investigated. The oxidation rates and activation energies were determined based on mass loss measurements in a series of oxidation tests. The surface morphology was characterized by scanning electron microscopy. Although the thermal oxidation mechanism was previously considered to be the same for all temperatures higher than 1000 °C, the significant increases in oxidation rate observed at very high temperatures suggest that the oxidation behavior of the selected graphite materials at temperatures higher than 1200 °C is different. This work demonstrates that changes in surface morphology and in oxidation rate of the filler particles in the graphite materials are more prominent at temperatures above 1200 °C. Furthermore, possible intrinsic factors contributing to the oxidation of the two graphite materials at different temperature ranges are discussed taking account of the dominant role played by temperature.

Chemical Characterization and Removal of Carbon-14 from Irradiated Graphite II - 13023

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

Dunzik-Gougar, Mary Lou; Cleaver, James; LaBrier, Daniel

2013-07-01

Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented last year and updated here is to identify the chemical form of C-14more » in irradiated graphite and develop a practical method by which C-14 can be removed. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam{sup R}, were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10{sup 13} neutrons/cm{sup 2}/s). Finer grained NBG-25 was not exposed to liquid nitrogen prior to irradiation at a neutron flux on the order of 10{sup 14} /cm{sup 2}/s. Characterization of pre- and post-irradiation graphite was conducted to determine the chemical environment and quantity of C-14 and its precursors via the use of surface sensitive characterization techniques. Scanning Electron Microscopy (SEM) was used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Energy Dispersive X-ray Analysis Spectroscopy (EDX). Results of post-irradiation characterization of these materials indicate a variety of surface functional groups containing carbon, oxygen, nitrogen and hydrogen. During thermal treatment, irradiated graphite samples are heated in the presence of an inert carrier gas (with or without oxidant gas), which carries off gaseous products released during treatment. Graphite gasification occurs via interaction with adsorbed oxygen complexes. Experiments in argon were performed at 900 deg. C and 1400 deg. C to evaluate the selective removal of C-14. Thermal treatment also was performed with the addition of 3 and 5 volume % oxygen at temperatures 700 deg. C and 1400 deg. C. Thermal treatment experiments were evaluated for the effective selective removal of C-14. Lower temperatures and oxygen levels correlated to more efficient C-14 removal. (authors)« less

NEUTRONIC REACTORS

DOEpatents

Wigner, E.P.

1960-11-22

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT).

PubMed

Payne, Liam; Heard, Peter J; Scott, Thomas B

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK's first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600-1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment.

A Study of the Oxidation Behaviour of Pile Grade A (PGA) Nuclear Graphite Using Thermogravimetric Analysis (TGA), Scanning Electron Microscopy (SEM) and X-Ray Tomography (XRT)

PubMed Central

Payne, Liam; Heard, Peter J.; Scott, Thomas B.

2015-01-01

Pile grade A (PGA) graphite was used as a material for moderating and reflecting neutrons in the UK’s first generation Magnox nuclear power reactors. As all but one of these reactors are now shut down there is a need to understand the residual state of the material prior to decommissioning of the cores, in particular the location and concentration of key radio-contaminants such as 14C. The oxidation behaviour of unirradiated PGA graphite was studied, in the temperature range 600–1050°C, in air and nitrogen using thermogravimetric analysis, scanning electron microscopy and X-ray tomography to investigate the possibility of using thermal degradation techniques to examine 14C distribution within irradiated material. The thermal decomposition of PGA graphite was observed to follow the three oxidation regimes historically identified by previous workers with limited, uniform oxidation at temperatures below 600°C and substantial, external oxidation at higher temperatures. This work demonstrates that the different oxidation regimes of PGA graphite could be developed into a methodology to characterise the distribution and concentration of 14C in irradiated graphite by thermal treatment. PMID:26575374

FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Davidson, J.K.

1963-11-19

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

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.

Method and system for producing hydrogen using sodium ion separation membranes

DOEpatents

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

2013-05-21

A method of producing hydrogen from sodium hydroxide and water is disclosed. The method comprises separating sodium from a first aqueous sodium hydroxide stream in a sodium ion separator, feeding the sodium produced in the sodium ion separator to a sodium reactor, reacting the sodium in the sodium reactor with water, and producing a second aqueous sodium hydroxide stream and hydrogen. The method may also comprise reusing the second aqueous sodium hydroxide stream by combining the second aqueous sodium hydroxide stream with the first aqueous sodium hydroxide stream. A system of producing hydrogen is also disclosed.

Lubricant-Coolant for Hot Working of Metals,

DTIC Science & Technology

includes calcium acetate, sodium acetate, and polyoxyethylated alkylphenol for added effectiveness, and that its composition includes (in wt. percentage...calcium acetate 5, sodium acetate 4, polyoxyethylated alkylphenol 0.1, graphite 5, and water up to 100. (Author)

Modified Graphene Oxide for Long Cycle Sodium-Ion Batteries

NASA Astrophysics Data System (ADS)

Shareef, Muhamed; Gunn, Harrison; Voigt, Victoria; Singh, Gurpreet

Hummer's process was modified to produce gram levels of 2-dimensional nanosheets of graphene oxide (GO) with varying degree of exfoliation and chemical functionalization. This was achieved by varying the weight ratios and reaction times of oxidizing agents used in the process. Based on Raman and Fourier transform infra red spectroscopy we show that potassium permanganate (KMnO4) is the key oxidizing agent while sodium nitrate (NaNO3) and sulfuric acid (H2SO4) play minor role during the exfoliation of graphite. Tested as working electrode in sodium-ion half-cell, the GO nanosheets produced using this optimized approach showed high rate capability and exceptionally high energy density of ~500 mAh/g for up to at least 100 cycles, which is among the highest reported for sodium/graphite electrodes. The average Coulombic efficiency was approximately 99 %. NSF Grant No. 1454151.

Neutronics and Transient Calculations for the Conversion of the Transient Reactor Rest Facility (TREAT)

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

Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Papadias, Dionissios D.

2015-01-01

The Transient Reactor Test Facility (TREAT) is a graphite-reflected, graphitemoderated, and air-cooled reactor fueled with 93.1% enriched UO2 particles dispersed in graphite, with a carbon-to-235U ratio of ~10000:1. TREAT was used to simulate accident conditions by subjecting fuel test samples placed at the center of the core to high energy transient pulses. The transient pulse production is based on the core’s selflimiting nature due to the negative reactivity feedback provided by the fuel graphite as the core temperature rises. The analysis of the conversion of TREAT to low enriched uranium (LEU) is currently underway. This paper presents the analytical methodsmore » used to calculate the transient performance of TREAT in terms of power pulse production and resulting peak core temperatures. The validation of the HEU neutronics TREAT model, the calculation of the temperature distribution and the temperature reactivity feedback as well as the number of fissions generated inside fuel test samples are discussed.« less

NUCLEAR REACTOR ELEMENT

DOEpatents

Sanz, M.C.; Scully, C.N.

1961-06-27

The patented fuel element is a hexagonal graphite body having an axial channel therethrough. The graphite is impregnated with uranium which is concentrated near the axial channel. Layers of tantalum nitride and tantalum carbide are disposed on the surface of the body confronting the channel.

AIR COOLED NEUTRONIC REACTOR

DOEpatents

Fermi, E.; Szilard, L.

1958-05-27

A nuclear reactor of the air-cooled, graphite moderated type is described. The active core consists of a cubicle mass of graphite, approximately 25 feet in each dimension, having horizontal channels of square cross section extending between two of the opposite faces, a plurality of cylindrical uranium slugs disposed in end to end abutting relationship within said channels providing a space in the channels through which air may be circulated, and a cadmium control rod extending within a channel provided in the moderator. Suitable shielding is provlded around the core, as are also provided a fuel element loading and discharge means, and a means to circulate air through the coolant channels through the fuel charels to cool the reactor.

Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition.

NASA Astrophysics Data System (ADS)

Sharma, Uttam; Chauhan, Sachin S.; Sharma, Jayshree; Sanyasi, A. K.; Ghosh, J.; Choudhary, K. K.; Ghosh, S. K.

2016-10-01

The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS.

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

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

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

Enhancing biodegradation and energy generation via roughened surface graphite electrode in microbial desalination cell.

PubMed

Ebrahimi, Atieh; Yousefi Kebria, Daryoush; Najafpour Darzi, Ghasem

2017-09-01

The microbial desalination cell (MDC) is known as a newly developed technology for water and wastewater treatment. In this study, desalination rate, organic matter removal and energy production in the reactors with and without desalination function were compared. Herein, a new design of plain graphite called roughened surface graphite (RSG) was used as the anode electrode in both microbial fuel cell (MFC) and MDC reactors for the first time. Among the three type of anode electrodes investigated in this study, RSG electrode produced the highest power density and salt removal rate of 10.81 W/m 3 and 77.6%, respectively. Such a power density was 2.33 times higher than the MFC reactor due to the junction potential effect. In addition, adding the desalination function to the MFC reactor enhanced columbic efficiency from 21.8 to 31.4%. These results provided a proof-of-concept that the use of MDC instead of MFC would improve wastewater treatment efficiency and power generation, with an added benefit of water desalination. Furthermore, RSG can successfully be employed in an MDC or MFC, enhancing the bio-electricity generation and salt removal.

Characterization of Sodium Thermal Hydraulics with Optical Fiber Temperature Sensors

NASA Astrophysics Data System (ADS)

Weathered, Matthew Thomas

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

Ab initio and Molecular Dynamic models of displacement damage in crystalline and turbostratic graphite

NASA Astrophysics Data System (ADS)

McKenna, Alice

One of the functions of graphite is as a moderator in several nuclear reactor designs, including the Advanced Gas-cooled Reactor (AGR). In the reactor graphite is used to thermalise the neutrons produced in the fission reaction thus allowing a self-sustained reaction to occur. The graphite blocks, acting as the moderator, are constantly irradiated and consequently suffer damage. This thesis examines the types of damage caused using molecular dynamic (MD) simulations and ab intio calculations. Neutron damage starts with a primary knock-on atom (PKA), which is travelling so fast that it creates damage through electronic and thermal excitation (this is addressed with thermal spike simulations). When the PKA has lost energy the subsequent cascade is based on ballistic atomic displacement. These two types of simulations were performed on single crystal graphite and other carbon structures such as diamond and amorphous carbon as a comparison. The thermal spike in single crystal graphite produced results which varied from no defects to a small number of permanent defects in the structure. It is only at the high energy range that more damage is seen but these energies are less likely to occur in the nuclear reactor. The thermal spike does not create damage but it is possible that it can heal damaged sections of the graphite, which can be demonstrated with the motion of the defects when a thermal spike is applied. The cascade simulations create more damage than the thermal spike even though less energy is applied to the system. A new damage function is found with a threshold region that varies with the square root of energy in excess of the energy threshold. This is further broken down in to contributions from primary and subsequent knock-on atoms. The threshold displacement energy (TDE) is found to be Ed=25eV at 300K. In both these types of simulation graphite acts very differently to the other carbon structures. There are two types of polycrystalline graphite structures which simulations have been performed on. The difference between the two is at the grain boundaries with one having dangling bonds and the other one being bonded. The cascade showed the grain boundaries acting as a trap for the knock-on atoms which produces more damage compared with the single crystal. Finally the effects of turbostratic disorder on damage is considered. Density functional theory (DFT) was used to look at interstitials in (002) twist boundaries and how they act compared to AB stacked graphite. The results of these calculations show that the spiro interstitial is more stable in these grain boundaries, so at temperatures where the interstitial can migrate along the c direction they will segregate to (002) twist boundaries.

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

IDENTIFICATION OF CHLOROMETHANE FROMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

EPA Science Inventory

The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...

IDENTIFICATION OF CHLOROMETHANE FORMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

EPA Science Inventory

The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...

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

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

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

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

DOE PAGES

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.; ...

2016-12-21

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

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

Qualls, A. Louis; Betzler, Benjamin R.; Brown, Nicholas R.

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor-II (EBR-II) for sodium-cooled fast reactors. Historically, engineering demonstrations have played a vital role in advancing the technology readiness level of reactor technologies. Our paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would usemore » tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Important capabilities to be demonstrated by building and operating the FHR DR include fabrication and operation of high temperature reactors; heat exchanger performance (including passive decay heat removal); pump performance; and reactivity control; salt chemistry control to maximize vessel life; tritium management; core design methodologies; salt procurement, handling, maintenance and ultimate disposal. It is recognized that non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.« less

EXPERIMENTAL LIQUID METAL FUEL REACTOR

DOEpatents

Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

1962-01-23

A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

JACKETED FUEL ELEMENTS FOR GRAPHITE MODERATED REACTORS

DOEpatents

Szilard, L.; Wigner, E.P.; Creutz, E.C.

1959-05-12

Fuel elements for a heterogeneous, fluid cooled, graphite moderated reactor are described. The fuel elements are comprised of a body of natural uranium hermetically sealed in a jacket of corrosion resistant material. The jacket, which may be aluminum or some other material which is non-fissionable and of a type having a low neutron capture cross-section, acts as a barrier between the fissioning isotope and the coolant or moderator or both. The jacket minimizes the tendency of the moderator and coolant to become radioactive and/or contaminated by fission fragments from the fissioning isotope.

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

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

Romano, T.

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is validmore » until October 1, 1999. After this date, an update or upgrade to this document is required.« less

7. Another picture of workers laying up the graphite core ...

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

7. Another picture of workers laying up the graphite core of the 105-B pile. This view is towards the rear of the pile. The gun barrels can be seen protruding into the pile. D-3047 - B Reactor, Richland, Benton County, WA

Evaluation of co-cokes from bituminous coal with vacuum resid or decant oil, and evaluation of anthracites, as precursors to graphite

NASA Astrophysics Data System (ADS)

Nyathi, Mhlwazi S.

2011-12-01

Graphite is utilized as a neutron moderator and structural component in some nuclear reactor designs. During the reactor operaction the structure of graphite is damaged by collision with fast neutrons. Graphite's resistance to this damage determines its lifetime in the reactor. On neutron irradiation, isotropic or near-isotropic graphite experiences less structural damage than anisotropic graphite. The degree of anisotropy in a graphite artifact is dependent on the structure of its precursor coke. Currently, there exist concerns over a short supply of traditional precursor coke, primarily due to a steadily increasing price of petroleum. The main goal of this study was to study the anisotropic and isotropic properties of graphitized co-cokes and anthracites as a way of investigating the possibility of synthesizing isotropic or near-isotropic graphite from co-cokes and anthracites. Demonstrating the ability to form isotropic or near-isotropic graphite would mean that co-cokes and anthracites have a potential use as filler material in the synthesis of nuclear graphite. The approach used to control the co-coke structure was to vary the reaction conditions. Co-cokes were produced by coking 4:1 blends of vacuum resid/coal and decant oil/coal at temperatures of 465 and 500 °C for reaction times of 12 and 18 hours under autogenous pressure. Co-cokes obtained were calcined at 1420 °C and graphitized at 3000 °C for 24 hours. Optical microscopy, X-ray diffraction, temperature-programmed oxidation and Raman spectroscopy were used to characterize the products. It was found that higher reaction temperature (500 °C) or shorter reaction time (12 hours) leads to an increase in co-coke structural disorder and an increase in the amount of mosaic carbon at the expense of textural components that are necessary for the formation of anisotropic structure, namely, domains and flow domains. Characterization of graphitized co-cokes showed that the quality, as expressed by the degree of graphitization and crystallite dimensions, of the final product is dependent on the nature of the precursor co-coke. The methodology for studying anthracites was to select two anthracites on basis of rank, PSOC1515 being semi-anthracite and DECS21 anthracite. The selected anthracites were graphitized, in both native and demineralized states, under the same conditions as co-cokes. Products obtained from DECS21 showed higher degrees of graphitization and larger crystallite dimensions than products obtained from PSOC1515. Demineralization of anthracites served to increase the degree of graphitization, indicating that the minerals contained in these anthracites have no graphitization-enhancing ability. A larger crystallite length for products obtained from native versions, compared to demineralized versions, was attributed to a formation and decomposition of a silicon carbide during graphitization of native versions. In order to examine the anisotropic and isotropic properties, nuclear-grade graphite samples obtained from Oak Ridge National Laboratory (ORNL) and commercial graphite purchased from Fluka were characterized under similar conditions as graphitized co-cokes and anthracites. These samples served as representatives of "two extremes", with ORNL samples being the isotropic end and commercial graphite being the anisotropic end. Through evaluating relationships between structural parameters, it was observed that graphitized co-cokes are situated, structurally, somewhere between the "two extremes", whereas graphitized anthracites are closer to the anisotropic end. Basically, co-cokes have a better potential than anthracites to transform to isotropic or near-isotropic graphite upon graphitization. By co-coking vacuum resid/coal instead of decant oil/coal or using 500 °C instead of 465 °C, a shift away from commercial graphite towards ORNL samples was attained. Graphitizing a semi-anthracite or demineralizing anthracites before graphitization also caused a shift towards ORNL samples.

Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

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

Ogawa, Masuro; Kunugi, Tomoaki; Seki, Yasushi

If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basicmore » phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. The authors plan a scaled-model test and a full-scale model test for the LOVE.« less

Internal graphite moderator forces study, C and K Reactors

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

Cooley, D.E.

1963-10-28

The purpose of this study was to determine the maximum forces that can be imposed by the graphite moderator on prospective VSR channel sleeves. In order to do this, both the origins and modes of transmission of the forces were determined. Forces in the moderator stack that are capable of acting on a block or group of blocks may originate from any of the following primary effects: Contraction of graphite due to irradiation; thermal expansion of graphite; frictional resistance to motion; resistance from keys; gravity; and other.

Effect of Reacting Surface Density on the Overall Graphite Oxidation Rate

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

Chang H. Oh; Eung Kim; Jong Lim

2009-05-01

Graphite oxidation in an air-ingress accident is presently a very important issue for the reactor safety of the very high temperature gas cooled-reactor (VHTR), the concept of the next generation nuclear plant (NGNP) because of its potential problems such as mechanical degradation of the supporting graphite in the lower plenum of the VHTR might lead to core collapse if the countermeasure is taken carefully. The oxidation process of graphite has known to be affected by various factors, including temperature, pressure, oxygen concentration, types of graphite, graphite shape and size, flow distribution, etc. However, our recent study reveals that the internalmore » pore characteristics play very important roles in the overall graphite oxidation rate. One of the main issues regarding graphite oxidation is the potential core collapse problem that may occur following the degradation of graphite mechanical strength. In analyzing this phenomenon, it is very important to understand the relationship between the degree of oxidization and strength degradation. In addition, the change of oxidation rate by graphite oxidation degree characterization by burn-off (ratio of the oxidized graphite density to the original density) should be quantified because graphite strength degradation is followed by graphite density decrease, which highly affects oxidation rates and patterns. Because the density change is proportional to the internal pore surface area, they should be quantified in advance. In order to understand the above issues, the following experiments were performed: (1)Experiment on the fracture of the oxidized graphite and validation of the previous correlations, (2) Experiment on the change of oxidation rate using graphite density and data collection, (3) Measure the BET surface area of the graphite. The experiments were performed using H451 (Great Lakes Carbon Corporation) and IG-110 (Toyo Tanso Co., Ltd) graphite. The reason for the use of those graphite materials is because their chemical and mechanical characteristics are well identified by the previous investigations, and therefore it was convenient for us to access the published data, and to apply and validate our new methodologies. This paper presents preliminary results of compressive strength vs. burn-off and surface area density vs. burn-off, which can be used for the nuclear graphite selection for the NGNP.« less

ICP-MS measurement of diffusion coefficients of Cs in NBG-18 graphite

NASA Astrophysics Data System (ADS)

Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2015-11-01

Graphite is used in the HGTR/VHTR as moderator and it also functions as a barrier to fission product release. Therefore, an elucidation of transport of fission products in reactor-grade graphite is required. We have measured diffusion coefficients of Cs in graphite NBG-18 using the release method, wherein we infused spheres of NBG-18 with Cs and measured the release rates in the temperature range of 1090-1395 K. We have obtained: These seem to be the first reported values of Cs diffusion coefficients in NBG-18. The values are lower than those reported for other graphites in the literature.

Neutronic reactor construction

DOEpatents

Huston, Norman E.

1976-07-06

1. A neutronic reactor comprising a moderator including horizontal layers formed of horizontal rows of graphite blocks, alternate layers of blocks having the rows extending in one direction, the remaining alternate layers having the rows extending transversely to the said one direction, alternate rows of blocks in one set of alternate layers having longitudinal ducts, the moderator further including slotted graphite tubes positioned in the ducts, the reactor further comprising an aluminum coolant tube positioned within the slotted tube in spaced relation thereto, bodies of thermal-neutron-fissionable material, and jackets enclosing the bodies and being formed of a corrosion-resistant material having a low neutron-capture cross section, the bodies and jackets being positioned within the coolant tube so that the jackets are spaced from the coolant tube.

Ion irradiation used as surrogate of neutron irradiation in graphite: Consequences on 14C and 36Cl behavior and structural evolution

NASA Astrophysics Data System (ADS)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2018-04-01

Graphite has been widely used as neutron moderator, reflector or fuel matrix in different types of reactors such as gas cooled nuclear reactors (UNGG, Magnox, AGR), RBMK reactors or high temperature gas cooled reactors. Their operation produces a great quantity of irradiated graphite or other carbonaceous waste (around 250,000 tons worldwide) that requires a special management strategy. In the case of disposal, which is a current management strategy, two main radionuclides, 14C and 36Cl might be dose determining at the outlet. Particular attention is paid to 14C due to its long half-life (T∼5730 years) [1] and as major contributor to the radioactive dose. 14C has two main production routes, i) transmutation of nitrogen (14N(n,p)14C) where nitrogen is mainly adsorbed at the surfaces of the irradiated graphite; ii) activation of carbon from the matrix (13C(n,γ)14C). According to leaching tests, it was shown that even if the quantity of 14C released in the solution is low (less than 1% of the initial inventory), around 30% is in the organic form that would be mobile in repository conditions [2,3]. 36Cl is mainly produced through the activation of 35Cl (35Cl(n,γ)36Cl) which is an impurity in nuclear graphite. Its activity is low but it might be highly mobile in clay host rocks. Thus, in order to make informed decisions about the best management process and to anticipate potential radionuclide dissemination during dismantling and in the repository, it is necessary to collect information on 14C and 36Cl location and speciation in graphite, after reactor closure. The goal of the present paper is therefore to use ion irradiation to simulate neutron irradiation and to evaluate the irradiation effects on the behavior of 36Cl and 14C as well as on the induced graphite structure modifications. For that, to understand and model the underlying mechanisms, we used an indirect approach based on 13C or 37Cl implantation to simulate the respective presence of 14C or 36Cl. These isotopes were implanted into Highly Oriented Pyrolytic Graphite (HOPG) samples used as a model material system representative of the nuclear graphite coke grains which form around 80% of nuclear graphite. Nuclear graphite is manufactured from petroleum coke grains (filler) blended with coal tar pitch acting as a binder. Shaped blocks are formed by extrusion of the blend. They are heat-treated up to about 2800 °C (graphitisation treatment) and polycrystalline graphite is obtained. Blocks, intended for the moderator or reflector, may be further impregnated with pitch, re-baked and regraphitised in order to increase the density. Virgin nuclear graphites have initial densities in the range 1.6-1.8 g cm-3. The difference with graphite crystal (density = 2.265 g cm-3) is due to internal porosity. As a result of mixing of several carbon compounds, this material is structurally heterogeneous at a local scale. Nuclear graphite presents a complex multiscale organisation. It can be locally more or less anisotropic and not completely graphitised. Nuclear graphite has a polycrystalline structure and contains micrometer sized grains. The grains are formed by several more or less oriented crystallites with a size of a few hundreds nanometers. Each crystallite is formed by a triperiodical stacking of graphene planes. Nuclear graphite contains also small amounts of impurities like oxygen, hydrogen, metals and halogens, among them chlorine [4]. Ion beam irradiation was used as a surrogate for neutrons because it may produce cascades (due to ballistic interactions) that could be similar to those created by neutrons in the nuclear reactor. Ion beam (or electron beam) irradiation has been used for many years to simulate neutron irradiation. It has advantages such as for example the possibility to vary the irradiation conditions and sometimes to carry out in situ observations. Moreover, depending on the ion nature and energy, it allows covering a broad range of the neutron recoil spectrum and the rate at which atoms are displaced can be increased in comparison to reactor conditions. Dose rates can thus be much higher than under neutron irradiation allowing for higher amounts of displacements per atoms (dpa) to be reached within some days instead of months or years. Moreover, because there is no sample activation, the samples are not radioactive [5-11]. During neutron irradiation, the neutrons interact with the matter both by collision with the atom nuclei (i.e. ballistic damage) and by nuclear reactions. The first atoms hit by neutrons are caused to move, thus starting a cascade of atomic collisions leading to electronic excitation as they go through the matter and on the path of the atoms they displace (recoil atoms). The ballistic damage can be evaluated using the nuclear stopping power and can be denoted by the number of displacements per atom (dpa). The effect of electronic excitation can be quantified using the electronic stopping power. The experimental simulation of neutron irradiation in a reactor can be done by irradiation of the graphite samples with different ions of different energies. The choice of these parameters enables the study of the damage effects with or without electron excitation or ballistic damage. Thus, knowing that the impinging neutrons induce mainly ballistic damage into the graphite matrix but that part of the recoil carbon energy is also transferred through electronic excitation, it is interesting to use ion irradiation because both ballistic damage and electronic excitation effects can be studied coupled or decoupled according to the nature of the ion, its energy and the fluence. It is possible to cover a wide range of electronic and nuclear stopping powers by working with different particle accelerators. Thus, we simulated the effects of these different irradiation regimes using ion irradiation by varying the Sn(nuclear)/Se(electronic) stopping power ratio as well as the irradiation temperature (from room temperature up to 1000 °C). Indeed, during reactor operation, neutron irradiation leads to changes in the graphite lattice parameters depending on irradiation conditions such as flux and fluence but also temperature [12]. Finally, Secondary Ion Mass Spectrometry (SIMS) analysis was used to determine 13C and 37Cl distribution profiles and allowed us to follow the implanted isotopes behavior. The structural modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry.

Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

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

Graf, A.; Petrick, H.; Stutz, U.

2012-07-01

The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up intomore » small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and also foreign decommissioning projects can provide solutions with were unknown before. (authors)« less

Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

NASA Astrophysics Data System (ADS)

Cisneros, Anselmo Tomas, Jr.

The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP and PEBBED for a high temperature gas cooled pebble bed reactor. Three parametric studies were performed for exploring the design space of the PB-FHR---to select a fuel design for the PB-FHR] to select a core configuration; and to optimize the PB-FHR design. These parametric studies investigated trends in the dependence of important reactor performance parameters such as burnup, temperature reactivity feedback, radiation damage, etc on the reactor design variables and attempted to understand the underlying reactor physics responsible for these trends. A pebble fuel parametric study determined that pebble fuel should be designed with a carbon to heavy metal ratio (C/HM) less than 400 to maintain negative coolant temperature reactivity coefficients. Seed and thorium blanket-, seed and inert pebble reflector- and seed only core configurations were investigated for annular FHR PBRs---the C/HM of the blanket pebbles and discharge burnup of the thorium blanket pebbles were additional design variable for core configurations with thorium blankets. Either a thorium blanket or graphite pebble reflector is required to shield the outer graphite reflector enough to extend its service lifetime to 60 EFPY. The fuel fabrication costs and long cycle lengths of the thorium blanket fuel limit the potential economic advantages of using a thorium blanket. Therefore, the seed and pebble reflector core configuration was adopted as the baseline core configuration. Multi-objective optimization with respect to economics was performed for the PB-FHR accounting for safety and other physical design constraints derived from the high-level safety regulatory criteria. These physical constraints were applied along in a design tool, Nuclear Application Value Estimator, that evaluated a simplified cash flow economics model based on estimates of reactor performance parameters calculated using correlations based on the results of parametric design studies for a specific PB-FHR design and a set of economic assumptions about the electricity market to evaluate the economic implications of design decisions. The optimal PB-FHR design---Mark 1 PB-FHR---is described along with a detailed summary of its performance characteristics including: the burnup, the burnup evolution, temperature reactivity coefficients, the power distribution, radiation damage distributions, control element worths, decay heat curves and tritium production rates. The Mk1 PB-FHR satisfies the PB-FHR safety criteria. The fuel, moderator (pebble core, pebble shell, graphite matrix, TRISO layers) and coolant have global negative temperature reactivity coefficients and the fuel temperatures are well within their limits.

Nuclear fuel elements and method of making same

DOEpatents

Schweitzer, Donald G.

1992-01-01

A nuclear fuel element for a high temperature gas nuclear reactor that has an average operating temperature in excess of 2000.degree. C., and a method of making such a fuel element. The fuel element is characterized by having fissionable fuel material localized and stabilized within pores of a carbon or graphite member by melting the fissionable material to cause it to chemically react with the carbon walls of the pores. The fissionable fuel material is further stabilized and localized within the pores of the graphite member by providing one or more coatings of pyrolytic carbon or diamond surrounding the porous graphite member so that each layer defines a successive barrier against migration of the fissionable fuel from the pores, and so that the outermost layer of pyrolytic carbon or diamond forms a barrier between the fissionable material and the moderating gases used in an associated high temperature gas reactor. The method of the invention provides for making such new elements either as generally spherically elements, or as flexible filaments, or as other relatively small-sized fuel elements that are particularly suited for use in high temperature gas reactors.

Plasma carburizing with surface micro-melting

NASA Astrophysics Data System (ADS)

Balanovsky, A. E.; Grechneva, M. V.; Van Huy, Vu; Ponomarev, B. B.

2018-03-01

This paper presents carburizing the surface of 20 low carbon steel using electric arc and graphite prior. A carbon black solution was prepared with graphite powder and sodium silicate in water. A detailed analysis of the phase structure and the distribution profile of the sample hardness after plasma treatment were given. The hardened layer consists of three different zones: 1 – the cemented layer (thin white zone) on the surface, 2 – heat-affected zone (darkly etching structure), 3 – the base metal. The experimental result shows that the various microstructures and micro-hardness profiles were produced depending on the type of graphite coating (percentage of liquid glass) and processing parameters. The experiment proved that the optimum content of liquid glass in graphite coating is 50–87.5%. If the amount of liquid glass is less than 50%, adhesion to metal is insufficient. If liquid glass content is more than 87.5%, carburization of a metal surface does not occur. A mixture of the eutectic lamellar structure, martensite and austenite was obtained by using graphite prior with 67% sodium silicate and the levels of the hardness layer increased to around 1000 HV. The thickness of the cemented layer formed on the surface was around 200 μm. It is hoped that this plasma surface carburizing treatment could improve the tribological resistance properties.

Plant and microorganisms support media for electricity generation in biological fuel cells with living hydrophytes.

PubMed

Salinas-Juárez, María Guadalupe; Roquero, Pedro; Durán-Domínguez-de-Bazúa, María Del Carmen

2016-12-01

Plant support media may impact power output in a biological fuel cell with living plants, due to the physical and biochemical processes that take place in it. A material for support medium should provide the suitable conditions for the robust microbial growth and its metabolic activity, degrading organic matter and other substances; and, transferring electrons to the anode. To consider the implementation of this type of bio-electrochemical systems in constructed wetlands, this study analyzes the electrochemical behavior of biological fuel cells with the vegetal species Phragmites australis, by using two different support media: graphite granules and a volcanic slag, commonly known as tezontle (stone as light as hair, from the Aztec or Nahuatl language). Derived from the results, both, graphite and tezontle have the potential to be used as support medium for plants and microorganisms supporting a maximum power of 26.78mW/m(2) in graphite reactors. These reactors worked under mixed control: with ohmic and kinetic resistances of the same order of magnitude. Tezontle reactors operated under kinetic control with a high activation resistance supplying 9.73mW/m(2). These performances could be improved with stronger bacterial populations in the reactor, to ensure the rapid depletion of substrate. Copyright © 2016 Elsevier B.V. All rights reserved.

Novel duplex vapor-electrochemical method for silicon solar cells. [reaction of fluorine and silicon compounds with sodium

NASA Technical Reports Server (NTRS)

Nanis, L.; Sanjurjo, A.; Sancier, K. M.; Bartlett, R.; Westphal, S.

1979-01-01

The dependence of the SiF4 Na reaction initiation time and of the efficiency of the reaction on Na particle size and reaction temperature were studied. Close to 100 percent utilization of Na was obtained, and formation of byproduct fluoro-silicate was decreased to below 10 percent. A SiF4 Na reactor was built to scale up the reaction by a factor of about four and is now being tested. A scaled up melting system was built and successfully used to separate Si from kilogram quantities of SiF4 NaF mixtures. Support studies of the volatilization of NaF performed in a smaller melting system indicated minimal loss of NaF as vapor at 1410 C. The wetting of graphite was also investigated to determine the constituents of the NaF phase which promote good wetting.

Sodium leak detection system for liquid metal cooled nuclear reactors

DOEpatents

Modarres, Dariush

1991-01-01

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

A probabilisitic based failure model for components fabricated from anisotropic graphite

NASA Astrophysics Data System (ADS)

Xiao, Chengfeng

The nuclear moderator for high temperature nuclear reactors are fabricated from graphite. During reactor operations graphite components are subjected to complex stress states arising from structural loads, thermal gradients, neutron irradiation damage, and seismic events. Graphite is a quasi-brittle material. Two aspects of nuclear grade graphite, i.e., material anisotropy and different behavior in tension and compression, are explicitly accounted for in this effort. Fracture mechanic methods are useful for metal alloys, but they are problematic for anisotropic materials with a microstructure that makes it difficult to identify a "critical" flaw. In fact cracking in a graphite core component does not necessarily result in the loss of integrity of a nuclear graphite core assembly. A phenomenological failure criterion that does not rely on flaw detection has been derived that accounts for the material behaviors mentioned. The probability of failure of components fabricated from graphite is governed by the scatter in strength. The design protocols being proposed by international code agencies recognize that design and analysis of reactor core components must be based upon probabilistic principles. The reliability models proposed herein for isotropic graphite and graphite that can be characterized as being transversely isotropic are another set of design tools for the next generation very high temperature reactors (VHTR) as well as molten salt reactors. The work begins with a review of phenomenologically based deterministic failure criteria. A number of this genre of failure models are compared with recent multiaxial nuclear grade failure data. Aspects in each are shown to be lacking. The basic behavior of different failure strengths in tension and compression is exhibited by failure models derived for concrete, but attempts to extend these concrete models to anisotropy were unsuccessful. The phenomenological models are directly dependent on stress invariants. A set of invariants, known as an integrity basis, was developed for a non-linear elastic constitutive model. This integrity basis allowed the non-linear constitutive model to exhibit different behavior in tension and compression and moreover, the integrity basis was amenable to being augmented and extended to anisotropic behavior. This integrity basis served as the starting point in developing both an isotropic reliability model and a reliability model for transversely isotropic materials. At the heart of the reliability models is a failure function very similar in nature to the yield functions found in classic plasticity theory. The failure function is derived and presented in the context of a multiaxial stress space. States of stress inside the failure envelope denote safe operating states. States of stress on or outside the failure envelope denote failure. The phenomenological strength parameters associated with the failure function are treated as random variables. There is a wealth of failure data in the literature that supports this notion. The mathematical integration of a joint probability density function that is dependent on the random strength variables over the safe operating domain defined by the failure function provides a way to compute the reliability of a state of stress in a graphite core component fabricated from graphite. The evaluation of the integral providing the reliability associated with an operational stress state can only be carried out using a numerical method. Monte Carlo simulation with importance sampling was selected to make these calculations. The derivation of the isotropic reliability model and the extension of the reliability model to anisotropy are provided in full detail. Model parameters are cast in terms of strength parameters that can (and have been) characterized by multiaxial failure tests. Comparisons of model predictions with failure data is made and a brief comparison is made to reliability predictions called for in the ASME Boiler and Pressure Vessel Code. Future work is identified that would provide further verification and augmentation of the numerical methods used to evaluate model predictions.

Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

NASA Astrophysics Data System (ADS)

Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

2017-07-01

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10-6 K-1 (α∥) and 6.15 × 10-6 K-1 (α⊥) at the temperature range of 25-700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.

Nuclear reactor shield including magnesium oxide

DOEpatents

Rouse, Carl A.; Simnad, Massoud T.

1981-01-01

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

Passive shut-down heat removal system

DOEpatents

Hundal, Rolv; Sharbaugh, John E.

1988-01-01

An improved shut-down heat removal system for a liquid metal nuclear reactor of the type having a vessel for holding hot and cold pools of liquid sodium is disclosed herein. Generally, the improved system comprises a redan or barrier within the reactor vessel which allows an auxiliary heat exchanger to become immersed in liquid sodium from the hot pool whenever the reactor pump fails to generate a metal-circulating pressure differential between the hot and cold pools of sodium. This redan also defines an alternative circulation path between the hot and cold pools of sodium in order to equilibrate the distribution of the decay heat from the reactor core. The invention may take the form of a redan or barrier that circumscribes the inner wall of the reactor vessel, thereby defining an annular space therebetween. In this embodiment, the bottom of the annular space communicates with the cold pool of sodium, and the auxiliary heat exchanger is placed in this annular space just above the drawn-down level that the liquid sodium assumes during normal operating conditions. Alternatively, the redan of the invention may include a pair of vertically oriented, concentrically disposed standpipes having a piston member disposed between them that operates somewhat like a pressure-sensitive valve. In both embodiments, the cessation of the pressure differential that is normally created by the reactor pump causes the auxiliary heat exchanger to be immersed in liquid sodium from the hot pool. Additionally, the redan in both embodiments forms a circulation flow path between the hot and cold pools so that the decay heat from the nuclear core is uniformly distributed within the vessel.

Neutron irradiation damage of nuclear graphite studied by high-resolution transmission electron microscopy and Raman spectroscopy

NASA Astrophysics Data System (ADS)

Krishna, R.; Jones, A. N.; McDermott, L.; Marsden, B. J.

2015-12-01

Nuclear graphite components are produced from polycrystalline artificial graphite manufacture from a binder and filler coke with approximately 20% porosity. During the operational lifetime, nuclear graphite moderator components are subjected to fast neutron irradiation which contributes to the change of material and physical properties such as thermal expansion co-efficient, young's modulus and dimensional change. These changes are directly driven by irradiation-induced changes to the crystal structure as reflected through the bulk microstructure. It is therefore of critical importance that these irradiation changes and there implication on component property changes are fully understood. This work examines a range of irradiated graphite samples removed from the British Experimental Pile Zero (BEPO) reactor; a low temperature, low fluence, air-cooled Materials Test Reactor which operated in the UK. Raman spectroscopy and high-resolution transmission electron microscopy (HRTEM) have been employed to characterise the effect of increased irradiation fluence on graphite microstructure and understand low temperature irradiation damage processes. HRTEM confirms the structural damage of the crystal lattice caused by irradiation attributed to a high number of defects generation with the accumulation of dislocation interactions at nano-scale range. Irradiation-induced crystal defects, lattice parameters and crystallite size compared to virgin nuclear graphite are characterised using selected area diffraction (SAD) patterns in TEM and Raman Spectroscopy. The consolidated 'D'peak in the Raman spectra confirms the formation of in-plane point defects and reflected as disordered regions in the lattice. The reduced intensity and broadened peaks of 'G' and 'D' in the Raman and HRTEM results confirm the appearance of turbulence and disordering of the basal planes whilst maintaining their coherent layered graphite structure.

Property changes of G347A graphite due to neutron irradiation

DOE PAGES

Campbell, Anne A.; Katoh, Yutai; Snead, Mary A.; ...

2016-08-18

A new, fine-grain nuclear graphite, grade G347A from Tokai Carbon Co., Ltd., has been irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to study the materials property changes that occur when exposed to neutron irradiation at temperatures of interest for Generation-IV nuclear reactor applications. Specimen temperatures ranged from 290°C to 800 °C with a maximum neutron fluence of 40 × 10 25 n/m 2 [E > 0.1 MeV] (~30dpa). Lastly, observed behaviors include: anisotropic behavior of dimensional change in an isotropic graphite, Young's modulus showing parabolic fluence dependence, electrical resistivity increasing at low fluence and additionalmore » increase at high fluence, thermal conductivity rapidly decreasing at low fluence followed by continued degradation, and a similar plateau value of the mean coefficient of thermal expansion for all irradiation temperatures.« less

Containment Sodium Chemistry Models in MELCOR.

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

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

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

Production of an impermeable composite of irradiated graphite and glass by hot isostatic pressing as a long term leach resistant waste form

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

Fachinger, Johannes; Muller, Walter; Marsat, Eric

2013-07-01

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades. Themore » most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions. Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout. Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout. FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or negligible porosity and a water impermeable structure. Structural analysis shows that the glass in the composite has replaced the pores in the graphite structure. The typical pore volume of a graphite material is in the range of 20 vol.%. Therefore no volume increase will occur in comparison with the former graphite material. This IGM material will allow the encapsulation of graphite with package densities larger than 1.5 ton per cubic meter. Therefore a huge volume saving can be achieved by such an alternative encapsulation method. Disposal performance is also enhanced since little or no leaching of radionuclides is observed due to the impermeability of the material NNL and FNAG have proved that IGM can be produced by hot isostatic pressing (HIP) which has several advantages for radioactive materials over the HVP process. - The sealed HIP container avoids the release of any radionuclides. - The outside of the waste package is not contaminated. - The HIP process time is shorter than the HVP process time. The isostatic press avoids anisotropic density distributions. - Simple filling of the HIP container has advantages over the filling of an axial die. (authors)« less

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.

Improving molten fluoride salt and Xe135 barrier property of nuclear graphite by phenolic resin impregnation process

NASA Astrophysics Data System (ADS)

He, Zhao; Lian, Pengfei; Song, Yan; Liu, Zhanjun; Song, Jinliang; Zhang, Junpeng; Feng, Jing; Yan, Xi; Guo, Quangui

2018-02-01

A densification process has been conducted on isostatic graphite (IG-110, TOYO TANSO CO., Ltd., Japan) by impregnating phenolic resin to get the densified isostatic graphite (D-IG-110) with pore diameter of nearly 11 nm specifically for molten salt reactor application. The microstructure, mechanical, thermophysical and other properties of graphite were systematically investigated and compared before and after the densification process. The molten fluoride salt and Xe135 penetration in the graphite were evaluated in a high-pressure reactor and a vacuum device, respectively. Results indicated that D-IG-110 exhibited improved properties including infiltration resistance to molten fluoride salt and Xe135 as compared to IG-110 due to its low porosity of 2.8%, the average pore diameter of 11 nm and even smaller open pores on the surface of the graphite. The fluoride salt infiltration amount of IG-110 was 13.5 wt% under 1.5 atm and tended to be saturated under 3 atm with the fluoride salt occupation of 14.8 wt%. As to the D-IG-110, no salts could be detected even up to 10 atm attempted loading. The helium diffusion coefficient of D-IG-110 was 6.92 × 10-8 cm2/s, significantly less than 1.21 × 10-2 cm2/s of IG-110. If these as-produced properties for impregnated D-IG-110 could be retained during MSR operation, the material could prove effective at inhibiting molten fluoride salt and Xe135 inventories in the graphite.

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

NASA Astrophysics Data System (ADS)

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

2009-07-01

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

Comparisons of sodium void and Doppler reactivities in large oxide and carbide LMFBRs

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

Su, S F

Sodium void and Doppler reactivities in two full scale (3000 MWth) LMFBRs are analyzed; one is fueled with UO/sub 2/ - PuO/sub 2/ and the other is fueled with UC - PuC. These two reactors are analyzed for beginning of life as well as for beginning and end of equilibrium cycle conditions, and the variations of these two safety parameters with burnup are explained. A series of comperative analyses of these two and several hypothetical reactors are carried out to determine how differences in fuel type, sodium content, and heavy metal concentration between an oxide and a carbide reactor affectmore » their sodium void and Doppler reactivities. The effect of the presence of conrol poison on sodium void reactivity is also addressed.« less

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

Carroll, Mark C.

High-purity graphite is the core structural material of choice in the Very High Temperature Reactor (VHTR) design, a graphite-moderated, helium-cooled configuration capable of producing thermal energy for power generation as well as process heat for industrial applications that require temperatures higher than the outlet temperatures of present nuclear reactors. The Baseline Graphite Characterization Program is establishing accurate as-manufactured mechanical and physical property distributions in nuclear-grade graphites by providing comprehensive data that captures the level of variation in measured values. In addition to providing a thorough comparison between these values in different graphite grades, the program is also carefully tracking individualmore » specimen source, position, and orientation information in order to provide comparisons both in specific properties and in the associated variability between different lots, different billets, and different positions from within a single billet. This report is a preliminary comparison between each of the grades of graphite that are considered “candidate” grades from four major international graphite producers. These particular grades (NBG-18, NBG-17, PCEA, IG-110, and 2114) are the major focus of the evaluations presently underway on irradiated graphite properties through the series of Advanced Graphite Creep (AGC) experiments. NBG-18, a medium-grain pitch coke graphite from SGL from which billets are formed via vibration molding, was the favored structural material in the pebble-bed configuration. NBG-17 graphite from SGL is essentially NBG-18 with the grain size reduced by a factor of two. PCEA, petroleum coke graphite from GrafTech with a similar grain size to NBG-17, is formed via an extrusion process and was initially considered the favored grade for the prismatic layout. IG-110 and 2114, from Toyo Tanso and Mersen (formerly Carbone Lorraine), respectively, are fine-grain grades produced via an isomolding process. An analysis of the comparison between each of these grades will include not only the differences in fundamental and statistically-significant individual strength levels, but also the differences in the overall variability in properties within each of the grades that will ultimately provide the basis for predicting in-service performance. The comparative performance of the different types of nuclear-grade graphites will naturally continue to evolve as thousands more specimens are fully characterized with regard to strength, physical properties, and thermal performance from the numerous grades of graphite being evaluated.« less

Lighting Studies for Fuelling Machine Deployed Visual Inspection Tool

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

Stoots, Carl; Griffith, George

2015-04-01

Under subcontract to James Fisher Nuclear, Ltd., INL has been reviewing advanced vision systems for inspection of graphite in high radiation, high temperature, and high pressure environments. INL has performed calculations and proof-of-principle measurements of optics and lighting techniques to be considered for visual inspection of graphite fuel channels in AGR reactors in UK.

Development of a trickle bed reactor of electro-Fenton process for wastewater treatment.

PubMed

Lei, Yangming; Liu, Hong; Shen, Zhemin; Wang, Wenhua

2013-10-15

To avoid electrolyte leakage and gas bubbles in the electro-Fenton (E-Fenton) reactors using a gas diffusion cathode, we developed a trickle bed cathode by coating a layer composed of carbon black and polytetrafluoroethylene (C-PTFE) onto graphite chips instead of carbon cloth. The trickle bed cathode was optimized by single-factor and orthogonal experiments, in which carbon black, PTFE, and a surfactant were considered as the determinant of the performance of graphite chips. In the reactor assembled by the trickle bed cathode, H2O2 was generated with a current of 0.3A and a current efficiency of 60%. This performance was attributed to the fine distribution of electrolyte and air, as well as the effective oxygen transfer from the gas phase to the electrolyte-cathode interface. In terms of H2O2 generation and current efficiency, the developed trickle bed reactor had a performance comparable to that of the conventional E-Fenton reactor using a gas diffusion cathode. Further, 123 mg L(-1) of reactive brilliant red X-3B in aqueous solution was decomposed in the optimized trickle bed reactor as E-Fenton reactor. The decolorization ratio reached 97% within 20 min, and the mineralization reached 87% within 3h. Copyright © 2013 Elsevier B.V. All rights reserved.

FUEL ELEMENTS FOR NEUTRONIC REACTORS

DOEpatents

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

1963-05-01

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

The radioactivity estimation of 14C and 3H in graphite waste samples of the KRR-2.

PubMed

Reyoung Kim, Hee

2013-09-01

The radioactivity of (14)C and (3)H in graphite samples from the dismantled Korea Research Reactor-2 (the KRR-2) site was analyzed by high-temperature oxidation and liquid scintillation counting, and the graphite waste was suggested to be disposed of as a low-level radioactive waste. The graphite samples were oxidized at a high temperature of 800 °C, and their counting rates were measured by using a liquid scintillation counter (LSC). The combustion ratio of the graphite was about 99% on the sample with a maximum weight of 1g. The recoveries from the combustion furnace were around 100% and 90% in (14)C and (3)H, respectively. The minimum detectable activity was 0.04-0.05 Bq/g for the (14)C and 0.13-0.15 Bq/g for the (3)H at the same background counting time. The activity of (14)C was higher than that of (3)H over all samples with the activity ratios of the (14)C to (3)H, (14)C/(3)H, being between 2.8 and 25. The dose calculation was carried out from its radioactivity analysis results. The dose estimation gave a higher annual dose than the domestic legal limit for a clearance. It was thought that the sampled graphite waste from the dismantled research reactor was not available for reuse or recycling and should be monitored as low-level radioactive waste. Copyright © 2013 Elsevier Ltd. All rights reserved.

Metal fires and their implications for advanced reactors.

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

Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean

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

A liquid-metal filling system for pumped primary loop space reactors

NASA Astrophysics Data System (ADS)

Crandall, D. L.; Reed, W. C.

Some concepts for the SP-100 space nuclear power reactor use liquid metal as the primary coolant in a pumped loop. Prior to filling ground engineering test articles or reactor systems, the liquid metal must be purified and circulated through the reactor primary system to remove contaminants. If not removed, these contaminants enhance corrosion and reduce reliability. A facility was designed and built to support Department of Energy Liquid Metal Fast Breeder Reactor tests conducted at the Idaho National Engineering Laboratory. This test program used liquid sodium to cool nuclear fuel in in-pile experiments; thus, a system was needed to store and purify sodium inventories and fill the experiment assemblies. This same system, with modifications and potential changeover to lithium or sodium-potassium (NaK), can be used in the Space Nuclear Power Reactor Program. This paper addresses the requirements, description, modifications, operation, and appropriateness of using this liquid-metal system to support the SP-100 space reactor program.

A brief History of Neutron Scattering at the Oak Ridge High Flux Isotope Reactor

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

Nagler, Stephen E; Mook Jr, Herbert A

2008-01-01

Neutron scattering at the Oak Ridge National Laboratory dates back to 1945 when Ernest Wollan installed a modified x-ray diffractometer on a beam port of the original graphite reactor. Subsequently, Wollan and Clifford Shull pioneered neutron diffraction and laid the foundation for an active neutron scattering effort that continued through the 1950s, using the Oak Ridge Research reactor after 1958, and, starting in 1966, the High Flux Isotope Reactor, or HFIR.

An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

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

Bucknor, M.; Farmer, M.; Grabaskas, D.

The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination ofmore » gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.« less

Study of evaporating the irradiated graphite in equilibrium low-temperature plasma

NASA Astrophysics Data System (ADS)

Bespala, E. V.; Novoselov, I. Yu.; Pavlyuk, A. O.; Kotlyarevskiy, S. G.

2018-01-01

The paper describes a problem of accumulation of irradiated graphite due to operation of uranium-graphite nuclear reactors. The main noncarbon contaminants that contribute to the overall activity of graphite elements are iso-topes 137Cs, 60Co, 90Sr, 36Cl, and 3H. A method was developed for processing of irradiated graphite ensuring the volu-metric decontamination of samples. The calculation results are presented for equilibrium composition of plasma-chemical reactions in systems "irradiated graphite-argon" and "irradiated graphite-helium" for a wide range of tem-peratures. The paper describes a developed mathematical model for the process of purification of a porous graphite surface treated by equilibrium low-temperature plasma. The simulation results are presented for the rate of sublimation of radioactive contaminants as a function of plasma temperature and plasma flow velocity when different plasma-forming gases are used. The extraction coefficient for the contaminant 137Cs from the outer side of graphite pores was calculated. The calculations demonstrated the advantages of using a lighter plasma forming gas, i.e., helium.

REACTOR

DOEpatents

Szilard, L.

1963-09-10

A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

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

Takeda, T.; Shimazu, Y.; Hibi, K.

2012-07-01

Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of thismore » project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)« less

Thermal Properties of G-348 Graphite

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

McEligot, Donald; Swank, W. David; Cottle, David L.

2016-05-01

Fundamental measurements have been obtained in the INL Graphite Characterization Laboratory to deduce the temperature dependence of thermal conductivity for G-348 isotropic graphite, which has been used by City College of New York in thermal experiments related to gas-cooled nuclear reactors. Measurements of thermal diffusivity, mass, volume and thermal expansion were converted to thermal conductivity in accordance with ASTM Standard Practice C781-08. Data are tabulated and a preliminary correlation for the thermal conductivity is presented as a function of temperature from laboratory temperature to 1000C.

Thermal Properties of G-348 Graphite

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

McEligot, Donald M.; Swank, W. David; Cottle, David L.

Fundamental measurements have been obtained in the INL Graphite Characterization Laboratory to deduce the temperature dependence of thermal conductivity for G-348 isotropic graphite, which has been used by City College of New York in thermal experiments related to gas-cooled nuclear reactors. Measurements of thermal diffusivity, mass, volume and thermal expansion were converted to thermal conductivity in accordance with ASTM Standard Practice C781-08 (R-2014). Data are tabulated and a preliminary correlation for the thermal conductivity is presented as a function of temperature from laboratory temperature to 1000C.

Design of Modern Reactors for Synthesis of Thermally Expanded Graphite.

PubMed

Strativnov, Eugene V

2015-12-01

One of the most progressive trends in the development of modern science and technology is the creation of energy-efficient technologies for the synthesis of nanomaterials. Nanolayered graphite (thermally exfoliated graphite) is one of the key important nanomaterials of carbon origin. Due to its unique properties (chemical and thermal stability, ability to form without a binder, elasticity, etc.), it can be used as an effective absorber of organic substances and a material for seal manufacturing for such important industries as gas transportation and automobile. Thermally expanded graphite is a promising material for the hydrogen and nuclear energy industries. The development of thermally expanded graphite production is resisted by high specific energy consumption during its manufacturing and by some technological difficulties. Therefore, the creation of energy-efficient technology for its production is very promising.

Analysis of fatty acids by graphite plate laser desorption/ionization time-of-flight mass spectrometry.

PubMed

Park, K H; Kim, H J

2001-01-01

Fatty acids obtained from triglycerides (trioelin, tripalmitin), foods (milk, corn oil), and phospholipids (phosphotidylcholine, phosphotidylserine, phosphatidic acid) upon alkaline hydrolysis were observed directly without derivatization by graphite plate laser desorption/ionization time-of-flight mass spectrometry (GPLDI-TOFMS). Mass-to-charge ratios predicted for sodium adducts of expected fatty acids (e.g. palmitic, oleic, linoleic and arachidonic acids) were observed without interference. Although at present no quantitation is possible, the graphite plate method enables a simple and rapid qualitative analysis of fatty acids. Copyright 2001 John Wiley & Sons, Ltd.

Baseline Concept Description of a Small Modular High Temperature Reactor

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

Hans Gougar

2014-05-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

Baseline Concept Description of a Small Modular High Temperature Reactor

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

Gougar, Hans D.

2014-10-01

The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNPmore » were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the Generation IV program and its specific R&D needs will be included in this report when appropriate for comparison. The distinguishing features of the HTGR are the refractory (TRISO) coated particle fuel, the low-power density, graphite-moderated core, and the high outlet temperature of the inert helium coolant. The low power density and fuel form effectively eliminate the possibility of core melt, even upon a complete loss of coolant pressure and flow. The graphite, which constitutes the bulk of the core volume and mass, provides a large thermal buffer that absorbs fission heat such that thermal transients occur over a timespan of hours or even days. As chemically-inert helium is already a gas, there is no coolant temperature or void feedback on the neutronics and no phase change or corrosion product that could degrade heat transfer. Furthermore, the particle coatings and interstitial graphite retain fission products such that the source terms at the plant boundary remain well below actionable levels under all anticipated nominal and off-normal operating conditions. These attributes enable the reactor to supply process heat to a collocated industrial plant with negligible risk of contamination and minimal dynamic coupling of the facilities (Figure 1). The exceptional retentive properties of coated particle fuel in a graphite matrix were first demonstrated in the DRAGON reactor, a European research facility that began operation in 1964.« less

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

Tommasi, J.; Ruggieri, J. M.; Lebrat, J. F.

The latest release (2.1) of the ERANOS code system, using JEF-2.2, JEFF-3.1 and ENDF/B-VI r8 multigroup cross-section libraries is currently being validated on fast reactor critical experiments at CEA-Cadarache (France). This paper briefly presents the library effect studies and the detailed best-estimate validation studies performed up to now as part of the validation process. The library effect studies are performed over a wide range of experimental configurations, using simple model and method options. They yield global trends about the shift from JEF-2.2 to JEFF-3.1 cross-section libraries, that can be related to individual sensitivities and cross-section changes. The more detailed, best-estimate,more » calculations have been performed up to now over three experimental configurations carried out in the MASURCA critical facility at CEA-Cadarache: two cores with a softened spectrum due to large amounts of graphite (MAS1A' and MAS1B), and a core representative of sodium-cooled fast reactors (CIRANO ZONA2A). Calculated values have been compared to measurements, and discrepancies analyzed in detail using perturbation theory. Values calculated with JEFF-3.1 were found to be within 3 standard deviations of the measured values, and at least of the same quality as the JEF-2.2 based results. (authors)« less

APPARATUS FOR DETECTING AND LOCATING PRESENCE OF FLUIDS

DOEpatents

Williamson, R.R.

1958-09-16

A system is described fur detecting water leaks in water-cooled neutronic reactors by utilizing an electrical hygrometer having a resistance element variable with the moisture content. The graphite blocks, forming the moderator in many types of reactors, coniain ducts in which helium gas is circulated. When a leak occurs in a coolant tube, the water will seep through the graphite until it oozes into one of the helium ducts, where it will be swept along with the helium into a system of pipes that connect each of the helium ducts. By inserting an electric hygrometer in each of these pipes and connecting it to an alarm system, the moisture content of the helium will cause a change in the electrical resistance of the hygrometer which will initiate a signal alarm indicating the presence and position of the leaky water tube in the reactor.

Pipe connector

DOEpatents

Sullivan, Thomas E.; Pardini, John A.

1978-01-01

A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated.

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

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

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

1983-06-01

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

Methanol metabolism and archaeal community changes in a bioelectrochemical anaerobic digestion sequencing batch reactor with copper-coated graphite cathode.

PubMed

Park, Jungyu; Lee, Beom; Shi, Peng; Kwon, Hyejeong; Jeong, Sang Mun; Jun, Hangbae

2018-07-01

In this study, the metabolism of methanol and changes in an archaeal community were examined in a bioelectrochemical anaerobic digestion sequencing batch reactor with a copper-coated graphite cathode (BEAD-SBR Cu ). Copper-coated graphite cathode produced methanol from food waste. The BEAD-SBR Cu showed higher methanol removal and methane production than those of the anaerobic digestion (AD)-SBR. The methane production and pH of the BEAD-SBR Cu were stable even under a high organic loading rate (OLR). The hydrogenotrophic methanogens increased from 32.2 to 60.0%, and the hydrogen-dependent methylotrophic methanogens increased from 19.5 to 37.7% in the bulk of BEAD-SBR Cu at high OLR. Where methanol was directly injected as a single substrate into the BEAD-SBR Cu , the main metabolism of methane production was hydrogenotrophic methanogenesis using carbon dioxide and hydrogen released by the oxidation of methanol on the anode through bioelectrochemical reactions. Copyright © 2018 Elsevier Ltd. All rights reserved.

Electrochemical Exfoliation of Graphite in Aqueous Sodium Halide Electrolytes toward Low Oxygen Content Graphene for Energy and Environmental Applications.

PubMed

Munuera, J M; Paredes, J I; Enterría, M; Pagán, A; Villar-Rodil, S; Pereira, M F R; Martins, J I; Figueiredo, J L; Cenis, J L; Martínez-Alonso, A; Tascón, J M D

2017-07-19

Graphene and graphene-based materials have shown great promise in many technological applications, but their large-scale production and processing by simple and cost-effective means still constitute significant issues in the path of their widespread implementation. Here, we investigate a straightforward method for the preparation of a ready-to-use and low oxygen content graphene material that is based on electrochemical (anodic) delamination of graphite in aqueous medium with sodium halides as the electrolyte. Contrary to previous conflicting reports on the ability of halide anions to act as efficient exfoliating electrolytes in electrochemical graphene exfoliation, we show that proper choice of both graphite electrode (e.g., graphite foil) and sodium halide concentration readily leads to the generation of large quantities of single-/few-layer graphene nanosheets possessing a degree of oxidation (O/C ratio down to ∼0.06) lower than that typical of anodically exfoliated graphenes obtained with commonly used electrolytes. The halide anions are thought to play a role in mitigating the oxidation of the graphene lattice during exfoliation, which is also discussed and rationalized. The as-exfoliated graphene materials exhibited a three-dimensional morphology that was suitable for their practical use without the need to resort to any kind of postproduction processing. When tested as dye adsorbents, they outperformed many previously reported graphene-based materials (e.g., they adsorbed ∼920 mg g -1 for methyl orange) and were useful sorbents for oils and nonpolar organic solvents. Supercapacitor cells assembled directly from the as-exfoliated products delivered energy and power density values (up to 15.3 Wh kg -1 and 3220 W kg -1 , respectively) competitive with those of many other graphene-based devices but with the additional advantage of extreme simplicity of preparation.

Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

NASA Astrophysics Data System (ADS)

Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

2018-01-01

ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical experiment yields that we could verify the effects of heterogeneity of propagation medium on waves in Liquid sodium.

Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

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

Afifah, Maryam, E-mail: maryam.afifah210692@gmail.com; Su’ud, Zaki; Miura, Ryosuke

2015-09-30

Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don’t need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design.more » The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.« less

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

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

Belles, Randy; Poore, III, Willis P.; Brown, Nicholas R.

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-basedmore » description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.« less

NUCLEAR REACTOR

DOEpatents

Starr, C.

1963-01-01

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Preparation, Characterization and Utilization of Electrodes Coated with Polymeric Networks Formed by Gamma Radiation Crosslinking.

DTIC Science & Technology

1987-04-01

polymers such as poly[ diallyl dimethyl ammonium chloride] , poly [vinylbenzyl trimethyl ammonium chloride], poly[styrene sulfonic acid , sodium salt] and...poly[acrylic acid ], which would ordinarily dissolve from the electrode surface in aqueous solution unless crosslinked into a network, and several...Irradiation on a Water-Soluble Polymer: DDAC 8 E. Electrochemistry of DDAC Networks on Platinum and Graphite 10 F. Poly [acrylic acid ] Films on Graphite

Electrochemical treatment of evaporated residue of soak liquor generated from leather industry.

PubMed

Boopathy, R; Sekaran, G

2013-09-15

The organic and suspended solids present in soak liquor, generated from leather industry, demands treatment. The soak liquor is being segregated and evaporated in solar evaporation pans/multiple effect evaporator due to non availability of viable technology for its treatment. The residue left behind in the pans/evaporator does not carry any reuse value and also faces disposal threat due to the presence of high concentration of sodium chloride, organic and bacterial impurities. In the present investigation, the aqueous evaporated residue of soak liquor (ERSL) was treated by electrochemical oxidation. Graphite/graphite and SS304/graphite systems were used in electrochemical oxidation of organics in ERSL. Among these, graphite/graphite system was found to be effective over SS304/graphite system. Hence, the optimised conditions for the electrochemical oxidation of organics in ERSL using graphite/graphite system was evaluated by response surface methodology (RSM). The mass transport coefficient (km) was calculated based on pseudo-first order rate kinetics for both the electrode systems (graphite/graphite and SS304/graphite). The thermodynamic properties illustrated the electrochemical oxidation was exothermic and non-spontaneous in nature. The calculated specific energy consumption at the optimum current density of 50 mA cm(-2) was 0.41 kWh m(-3) for the removal of COD and 2.57 kWh m(-3) for the removal of TKN. Copyright © 2013 Elsevier B.V. All rights reserved.

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.

Decay heat of sodium fast reactor: Comparison of experimental measurements on the PHENIX reactor with calculations performed with the French DARWIN package

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

Benoit, J. C.; Bourdot, P.; Eschbach, R.

2012-07-01

A Decay Heat (DH) experiment on the whole core of the French Sodium-Cooled Fast Reactor PHENIX has been conducted in May 2008. The measurements began an hour and a half after the shutdown of the reactor and lasted twelve days. It is one of the experiments used for the experimental validation of the depletion code DARWIN thereby confirming the excellent performance of the aforementioned code. Discrepancies between measured and calculated decay heat do not exceed 8%. (authors)

ETR BUILDING, TRA642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. LIQUID SODIUM PIPING INSIDE CUBICLE SHOWN IN ID-33-G-101. INL NEGATIVE NO. HD24-3-4. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Adsorption and Electronic Structure of Sr and Ag Atoms on Graphite Surfaces: a First-Principles Study

NASA Astrophysics Data System (ADS)

Luo, Xiao-Feng; Fang, Chao; Li, Xin; Lai, Wen-Sheng; Sun, Li-Feng; Liang, Tong-Xiang

2013-06-01

The adsorption behaviors of radioactive strontium and silver nuclides on the graphite surface in a high-temperature gas-cooled reactor are studied by first-principles theory using generalized gradient approximation (GGA) and local density approximation (LDA) pseudo-potentials. It turns out that Sr prefers to be absorbed at the hollow of the carbon hexagonal cell by 0.54 eV (GGA), while Ag likes to sit right above the carbon atom with an adsorption energy of almost zero (GGA) and 0.45 eV (LDA). Electronic structure analysis reveals that Sr donates its partial electrons of the 4p and 5s states to the graphite substrate, while Ag on graphite is a physical adsorption without any electron transfer.

Graphite Waste Tank Cleanup and Decontamination under the Marcoule UP1 D and D Program - 13166

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

Thomasset, Philippe; Chabeuf, Jean-Michel; Thiebaut, Valerie

2013-07-01

The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pits had to be retrieved and transferred into a newer and safer pit. After an extensive R and D program, the equipment and process necessary for retrieval operations were designed, built and tested. Themore » innovative process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. The retrieval of the most irradiating fragments was a challenge. Today, the first pit is totally empty and its stainless steel walls have been decontaminated using gels. In the second pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. After more than 6 years of operations in sludge retrieval, a lot of experience was acquired from which important 'lessons learned' could be shared. (authors)« less

In-situ method for treating residual sodium

DOEpatents

Sherman, Steven R.; Henslee, S. Paul

2005-07-19

A unique process for deactivating residual sodium in Liquid Metal Fast Breeder Reactor (LMFBR) systems which uses humidified (but not saturated) carbon dioxide at ambient temperature and pressure to convert residual sodium into solid sodium bicarbonate.

In-Situ Method for Treating Residual Sodium

DOEpatents

Sherman, Steven R.; Henslee, S. Paul

2005-07-19

A unique process for deactivating residual sodium in Liquid Metal Fast Breeder Reactor (LMFBR) systems which uses humidified (but not saturated) carbon dioxide at ambient temperature and pressure to convert residual sodium into solid sodium bicarbonate.

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)

Aghina, L.O.B.; Bellini, I.W.

The main works and modifications performed on the Argonaut reactor of the Instituto de Engenharia Nuclear after 4 years of operation are described. New positioning and holding system for the fuel elements and graphite wedges for the reactor brought appreciable stability to the operation. The removal of the plastic layer, printing and inspection of the corrosion of the elements plater are also described. (INIS)

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 2: Application to EBR-II Primary Sodium System and Related Systems

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

Steven R. Sherman; Collin J. Knight

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decontamination and decomissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidifed carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, USA. This report is Part 2 of a two-part report. This second report provides a supplement to the first report and describes the application of the humdidified carbon dioxide technique ("carbonation") to the EBR-II primary tank, primary cover gas systems, and the intermediate heat exchanger. Future treatment plans are also provided.« less

Experimental validation of an 8 element EMAT phased array probe for longitudinal wave generation

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

Le Bourdais, Florian, E-mail: florian.lebourdais@cea.fr; Marchand, Benoit, E-mail: florian.lebourdais@cea.fr

2015-03-31

Sodium cooled Fast Reactors (SFR) use liquid sodium as a coolant. Liquid sodium being opaque, optical techniques cannot be applied to reactor vessel inspection. This makes it necessary to develop alternative ways of assessing the state of the structures immersed in the medium. Ultrasonic pressure waves are well suited for inspection tasks in this environment, especially using pulsed electromagnetic acoustic transducers (EMAT) that generate the ultrasound directly in the liquid sodium. The work carried out at CEA LIST is aimed at developing phased array EMAT probes conditioned for reactor use. The present work focuses on the experimental validation of amore » newly manufactured 8 element probe which was designed for beam forming imaging in a liquid sodium environment. A parametric study is carried out to determine the optimal setup of the magnetic assembly used in this probe. First laboratory tests on an aluminium block show that the probe has the required beam steering capabilities.« less

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

DOEpatents

Hunsbedt, A.; Boardman, C.E.

1995-04-11

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

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

DOEpatents

Hunsbedt, Anstein; Boardman, Charles E.

1995-01-01

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

SODIUM DEUTERIUM REACTOR

DOEpatents

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

1963-02-26

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

AGR-2 and AGR-3/4 Release-to-Birth Ratio Data Analysis

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

Pham, Binh T.; Einerson, Jeffrey J.; Scates, Dawn M.

A series of Advanced Gas Reactor (AGR) irradiation tests is being conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) low enriched fuel used in the High Temperature Gas-cooled Reactor (HTGR). Each AGR test consists of multiple independently controlled and monitored capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples embedded in the graphite enabling temperature control. AGR configuration and irradiation conditions are based on prismatic HTGR technology that is distinguished primarily through use of heliummore » coolant, a low-power-density ceramic core capable of withstanding very high temperatures, and TRISO coated particle fuel. Thus, these tests provide valuable irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, and support development and validation of fuel performance and fission product transport models and codes.« less

Hydrogen Storage in Diamond Powder Utilizing Plasma NaF Surface Treatment for Fuel Cell Applications

NASA Astrophysics Data System (ADS)

Leal, David A.; Velez, Angel; Prelas, Mark A.; Gosh, Tushar; Leal-Quiros, E.

2006-12-01

Hydrogen Fuel Cells offer the vital solution to the world's socio-political dependence on oil. Due to existing difficulty in safe and efficient hydrogen storage for fuel cells, storing the hydrogen in hydrocarbon compounds such as artificial diamond is a realistic solution. By treating the surface of the diamond powder with a Sodium Fluoride plasma exposure, the surface of the diamond is cleaned of unwanted molecules. Due to fluorine's electro negativity, the diamond powder is activated and ready for hydrogen absorption. These diamond powder pellets are then placed on a graphite platform that is heated by conduction in a high voltage circuit made of tungsten wire. Then, the injection of hydrogen gas into chamber allows the storage of the Hydrogen on the surface of the diamond powder. By neutron bombardment in the nuclear reactor, or Prompt Gamma Neutron Activation Analysis, the samples are examined for parts per million amounts of hydrogen in the sample. Sodium Fluoride surface treatment allows for higher mass percentage of stored hydrogen in a reliable, resistant structure, such as diamond for fuel cells and permanently alters the diamonds terminal bonds for re-use in the effective storage of hydrogen. The highest stored amount utilizing the NaF plasma surface treatment was 22229 parts per million of hydrogen in the diamond powder which amounts to 2.2229% mass increase.

Ultra-High Temperature ContinuousReactors based on Electro-thermal FluidizedBed Concept

DOE PAGES

Fedorov, Sergiy S.; Rohatgi, Upendra Singh; Barsukov, Igor V.; ...

2015-12-08

This paper presents the results of research and development in high-temperature (i.e. 2,000- 3,000ºС) continuous furnaces operating on the principle of electro-thermal fluidized bed for the purification of recycled, finely sized carbon materials. The basis of this fluidized bed furnace is specific electrical resistance and a new correlation has been developed to predict specific electrical resistance for the natural graphite-based precursors entering the fluidized bed reactor This correlation has been validated with the data from a fully functional pilot furnace whose throughput capacity is 10 kg per hour built as part of this work. Data collected in the course ofmore » graphite refining experiments demonstrated that difference between the calculated and measured values of specific electrical resistance of fluidized bed does not exceed 25%. It was concluded that due to chaotic nature of electro-thermal fluidized bed reactors this discrepancy is acceptable. The fluid mechanics of the three types of operating regimes, have been described. The numerical relationships obtained as part of this work allowed proposing an algorithm for selection of technological operational modes with large- scale high-temperature furnaces rated for throughputs of several tons of product per hour. Optimizations proposed now allow producing natural graphite-based end product with the purity level of 99.98+ wt%C which is the key passing criteria for applications in the advanced battery markets.« less

Design test request No. 1263 K Reactor graphite key and VSR channel sleeve test

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

Kempf, F.J.

1964-12-10

The objectives of this test were: (1) Determine the coefficient of friction between two adjacent layers of K Reactor graphite at room temperature. (2) Determine the average load required to cause failure of an unirradiated K Reactor side reflector bar, when subjected to tensile loading applied through the reflector keys. (3) Determine the average load at failure and the average deflection at failure of a single VSR channel key when loaded in keyways with clearances equal to those used in original stack construction. (4) Determine the average load and deflection required to break the four K Reactor VSR keys whenmore » loaded simultaneously in both `3-layer` and `7-layer` mockups. Also determine the mode of key failure; i.e., shear, flexure or combined compression and bending. Following these key rupture tests, determine the strength and deflection characteristics of the proposed K Reactor VSR channel sleeve when loaded in a manner identical to that used to fracture the keys. (5) Determine the average load and deflection at failure of both the proposed K Reactor VSR channel sleeves and the proposed C Reactor sleeves when subjected to crushing loads. (6) Determine the extent of damage to the proposed K Reactor VSR channel sleeve when subjected to the following vertical rod loading conditions. (a) Full rod drop in a channel mockup which has been misaligned 2 1/2 inches. (b) Full rod drop in a channel which has been misaligned an amount equal to the maximum flexibility of a `universal` VSR.« less

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

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

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

Influence of cyclic thermal loading on brazed composites for fusion applications

NASA Astrophysics Data System (ADS)

Šmid, I.; Kny, E.; Kneringer, G.; Reheis, N.

1990-04-01

Reactor grade graphite and molybdenum (TZM) were brazed with different high temperature brazes. The resulting tiles had a size of 50 × 50 mm2 with a graphite thickness of 10 mm and a TZM thickness of 5 mm. The brazed composites have been tested in electron beam simulation for their thermal fatigue properties. The parameters of these tests were chosen to match NET design specifications for normal operation and "slow" peak energy deposition. The resulting damage and microstructural changes on the graphites and the brazes are discussed. Additional information is supplied on X-ray diffraction data proving the presence of different phases in the brazes.

Diffusion of cesium and iodine in compressed IG-110 graphite compacts

NASA Astrophysics Data System (ADS)

Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2016-08-01

Nuclear graphite grade IG-110 is currently used in the High Temperature Engineering Test Reactor (HTTR) in Japan for certain permanent and replaceable core components, and is a material of interest in general. Therefore, transport parameters for fission products in this material are needed. Measurement of diffusion through pressed compacts of IG-110 graphite is experimentally attractive because they are easy to prepare with homogeneous distributions of fission product surrogates. In this work, we measured diffusion coefficients for Cs and I in pressed compacts made from IG-110 powder in the 1079-1290 K temperature range, and compared them to those obtained in as-received IG-110.

Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

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

Grabaskas, D.; Brunett, A.; Passerini, S.

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. Themore » mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.« less

Thermal Stratification Analysis for Sodium Fast Reactors

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

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

The WSTIAC Quarterly. Volume 9, Number 3

DTIC Science & Technology

2010-01-25

program .[8] THE THORIUM FUEL CYCLE AND LFTR POWER PLANT The thorium fuel cycle is based on a series of neutron absorp- tion and beta decay processes...the fig- ure is a graphite matrix moderated MSR reactor with fuel salt mixture (ThF4-U233F4) being circulated by a pump through the core and to a...the core as purified salt. As one of the unique safety features, a melt-plug at the reactor bottom would permit the reactor fluid fuel to be drained

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA603. SUMMARY OF ...

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

WATER PROCESS SYSTEM FLOW DIAGRAM FOR MTR, TRA-603. SUMMARY OF COOLANT FLOW FROM WORKING RESERVOIR TO INTERIOR OF REACTOR'S THERMAL SHIELD. NAMES TANK SECTIONS. PIPE AND DRAIN-LINE SIZES. SHOWS DIRECTION OF AIR FLOW THROUGH PEBBLE AND GRAPHITE BLOCK ZONE. NEUTRON CURTAIN AND THERMAL COLUMN DOOR. BLAW-KNOX 3150-92-7, 3/1950. INL INDEX NO. 531-0603-51-098-100036, REV. 6. - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

GAS COOLED NUCLEAR REACTORS

DOEpatents

Long, E.; Rodwell, W.

1958-06-10

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

Microfluidization of Graphite and Formulation of Graphene-Based Conductive Inks

PubMed Central

2017-01-01

We report the exfoliation of graphite in aqueous solutions under high shear rate [∼ 108 s–1] turbulent flow conditions, with a 100% exfoliation yield. The material is stabilized without centrifugation at concentrations up to 100 g/L using carboxymethylcellulose sodium salt to formulate conductive printable inks. The sheet resistance of blade coated films is below ∼2Ω/□. This is a simple and scalable production route for conductive inks for large-area printing in flexible electronics. PMID:28102670

The Development of an Electrostatically Safe Lead Styphnate Composition and the Reproduction of Lead Styphnate Mixtures by Wet Processes

DTIC Science & Technology

1946-10-01

possible replacements of gelatin:- H) ivie-,hyl cellulose (2) Sodium carboxy-methyl cellulose (3) AlbaJin (4) Casein (5) Gum arabio (6) Rei...there is no adequate explanation as to why some proteins such as casein and albuaen are unsatisfactory for use in the graphiting process; they have...stage, the continued addition of gelatin sensitises the excess graphite sol. so that the charge on the micelle is reduced to such an extent that

Method and apparatus for removing iodine from a nuclear reactor coolant

DOEpatents

Cooper, Martin H.

1980-01-01

A method and apparatus for removing iodine-131 and iodine-125 from a liquid sodium reactor coolant. Non-radioactive iodine is dissolved in hot liquid sodium to increase the total iodine concentration. Subsequent precipitation of the iodine in a cold trap removes both the radioactive iodine isotopes as well as the non-radioactive iodine.

Thermal oxidation of nuclear graphite: A large scale waste treatment option.

PubMed

Theodosiou, Alex; Jones, Abbie N; Marsden, Barry J

2017-01-01

This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF). Particulate samples of Magnox Reactor Pile Grade-A (PGA) graphite, were oxidised in both air and 60% O2, over the temperature range 400-1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700-800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000-1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput.

Thermal oxidation of nuclear graphite: A large scale waste treatment option

PubMed Central

Jones, Abbie N.; Marsden, Barry J.

2017-01-01

This study has investigated the laboratory scale thermal oxidation of nuclear graphite, as a proof-of-concept for the treatment and decommissioning of reactor cores on a larger industrial scale. If showed to be effective, this technology could have promising international significance with a considerable impact on the nuclear waste management problem currently facing many countries worldwide. The use of thermal treatment of such graphite waste is seen as advantageous since it will decouple the need for an operational Geological Disposal Facility (GDF). Particulate samples of Magnox Reactor Pile Grade-A (PGA) graphite, were oxidised in both air and 60% O2, over the temperature range 400–1200°C. Oxidation rates were found to increase with temperature, with a particular rise between 700–800°C, suggesting a change in oxidation mechanism. A second increase in oxidation rate was observed between 1000–1200°C and was found to correspond to a large increase in the CO/CO2 ratio, as confirmed through gas analysis. Increasing the oxidant flow rate gave a linear increase in oxidation rate, up to a certain point, and maximum rates of 23.3 and 69.6 mg / min for air and 60% O2 respectively were achieved at a flow of 250 ml / min and temperature of 1000°C. These promising results show that large-scale thermal treatment could be a potential option for the decommissioning of graphite cores, although the design of the plant would need careful consideration in order to achieve optimum efficiency and throughput. PMID:28793326

Pile construction

DOEpatents

Johnson, Alfred A.; Carleton, John T.

1978-05-02

A graphite-moderated, water-cooled nuclear reactor including graphite blocks disposed in transverse alternate layers, one set of alternate layers consisting of alternate full size blocks and smaller blocks through which cooling tubes containing fuel extend, said smaller blocks consisting alternately of tube bearing blocks and support block, the support blocks being smaller than the tube bearing blocks, the aperture of each support block being tapered so as to provide the tube extending therethrough with a narrow region of support while being elsewhere spaced therefrom.

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

Loyalka, Sudarshan

High and Very High Temperatures Gas Reactors (HTGRs/VHTRs) have five barriers to fission product (FP) release: the TRISO fuel coating, the fuel elements, the core graphite, the primary coolant system, and the reactor building. This project focused on measurements and computations of FP diffusion in graphite, FP adsorption on graphite and FP interactions with dust particles of arbitrary shape. Diffusion Coefficients of Cs and Iodine in two nuclear graphite were obtained by the release method and use of Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS) and Instrumented Neutron Activation Analysis (INAA). A new mathematical model for fission gas release from nuclear fuelmore » was also developed. Several techniques were explored to measure adsorption isotherms, notably a Knudsen Effusion Mass Spectrometer (KEMS) and Instrumented Neutron Activation Analysis (INAA). Some of these measurements are still in progress. The results will be reported in a supplemental report later. Studies of FP interactions with dust and shape factors for both chain-like particles and agglomerates over a wide size range were obtained through solutions of the diffusion and transport equations. The Green's Function Method for diffusion and Monte Carlo technique for transport were used, and it was found that the shape factors are sensitive to the particle arrangements, and that diffusion and transport of FPs can be hindered. Several journal articles relating to the above work have been published, and more are in submission and preparation.« less

DESIGN CHARACTERISTICS OF THE IDAHO NATIONAL LABORATORY HIGH-[TEMPERATURE GAS-COOLED TEST REACTOR

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

Sterbentz, James; Bayless, Paul; Strydom, Gerhard

A point design for a graphite-moderated, high-temperature, gas-cooled test reactor (HTG TR) has been developed by Idaho National Laboratory (INL) as part of a United States (U.S.) Department of Energy (DOE) initiative to explore and potentially expand the existing U.S. test reactor capabilities. This paper provides a summary of the design and its main attributes. The 200 MW HTG TR is a thermal-neutron spectrum reactor composed of hexagonal prismatic fuel and graphite reflector blocks. Twelve fuel columns (96 fuel blocks total and 6.34 m active core height) are arranged in two hexagonal rings to form a relatively compact, high-power density,more » annular core sandwiched between inner, outer, top, and bottom graphite reflectors. The HTG-TR is designed to operate at 7 MPa with a coolant inlet/outlet temperature of 325°C/650°C, and utilizes TRISO particle fuel from the DOE AGR Program with 425 ?m uranium oxycarbide (UCO) kernels and an enrichment of 15.5 wt% 235U. The primary mission of the HTG TR is material irradiation and therefore the core has been specifically designed and optimized to provide the highest possible thermal and fast neutron fluxes. The highest thermal neutron flux (3.90E+14 n/cm2s) occurs in the outer reflector, and the maximum fast flux levels (1.17E+14 n/cm2s) are produced in the central reflector column where most of the graphite has been removed. Due to high core temperatures under accident conditions, all the irradiation test facilities have been located in the inner and outer reflectors where fast flux levels decline. The core features a large number of irradiation positions with large test volumes and long test lengths, ideal for thermal neutron irradiation of large test articles. The total available test volume is more than 1100 liters. Up to four test loop facilities can be accommodated with pressure tube boundaries to isolate test articles and test fluids (e.g., liquid metal, liquid salt, light water) from the helium primary coolant system.« less

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

Danial, Wan Hazman, E-mail: hazmandanial@gmail.com; Majid, Zaiton Abdul, E-mail: zaiton@kimia.fs.utm.my; Aziz, Madzlan

The present work reports the synthesis and characterization of graphene via electrochemical exfoliation of graphite rod using two-electrode system assisted by Sodium Dodecyl Sulphate (SDS) as a surfactant. The electrochemical process was carried out with sequence of intercalation of SDS onto the graphite anode followed by exfoliation of the SDS-intercalated graphite electrode when the anode was treated as cathode. The effect of intercalation potential from 5 V to 9 V and concentration of the SDS surfactant of 0.1 M and 0.01 M were investigated. UV-vis Spectroscopic analysis indicated an increase in the graphene production with higher intercalation potential. Transmission Electron Microscopy (TEM)more » analysis showed a well-ordered hexagonal lattice of graphene image and indicated an angle of 60° between two zigzag directions within the honeycomb crystal lattice. Raman spectroscopy analysis shows the graphitic information effects after the exfoliation process.« less

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

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

Steven R. Sherman

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/ormore » to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium treatment within the EBR-II primary sodium cooling system and related systems.« less

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

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

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

A facile solvothermal method to produce graphene-ZnS composites for superior photoelectric applications

NASA Astrophysics Data System (ADS)

Lei, Yun; Chen, Feifei; Li, Rong; Xu, Jun

2014-07-01

In this experiment, flake graphite (<30 μm) was prepared as raw materials. Graphite oxide is prepared with Hummers method by low temperature, middle temperature and high temperature, and further treated with super-sonic oscillation to get graphene oxide. Graphene-zinc sulfide composites were synthesized through a simple solvothermal method using thiourea or sodium sulfide as sulfur source in the ethylene glycol or ethylenediamine, respectively. The products were characterized by X-ray and SEM, and analyzed by the transient photocurrent response and electrochemical impedance spectra. The results indicate that the properties of graphene-zinc sulfide composites prepared with thiourea in ethylene glycol are superior to those of blank-ZnS and composites prepared with sodium sulfide and ethylenediamine, which is attributed to electron capture and transfer ability of graphene resulting in a more efficient separation of the photoexcited charge carriers from ZnS-graphene composites.

Method of making a current collector for a sodium/sulfur battery

DOEpatents

Tischer, R.P.; Winterbottom, W.L.; Wroblowa, H.S.

1987-03-10

This specification is directed to a method of making a current collector for a sodium/sulfur battery. The current collector so-made is electronically conductive and resistant to corrosive attack by sulfur/polysulfide melts. The method includes the step of forming the current collector for the sodium/sulfur battery from a composite material formed of aluminum filled with electronically conductive fibers selected from the group of fibers consisting essentially of graphite fibers having a diameter up to 10 microns and silicon carbide fibers having a diameter in a range of 500--1,000 angstroms. 2 figs.

Method of making a current collector for a sodium/sulfur battery

DOEpatents

Tischer, Ragnar P.; Winterbottom, Walter L.; Wroblowa, Halina S.

1987-01-01

This specification is directed to a method of making a current collector (14) for a sodium/sulfur battery (10). The current collector so-made is electronically conductive and resistant to corrosive attack by sulfur/polysulfide melts. The method includes the step of forming the current collector for the sodium/sulfur battery from a composite material (16) formed of aluminum filled with electronically conductive fibers selected from the group of fibers consisting essentially of graphite fibers having a diameter up to 10 microns and silicon carbide fibers having a diameter in a range of 500-1000 angstroms.

Heat and mass transfer rates during flow of dissociated hydrogen gas over graphite surface

NASA Technical Reports Server (NTRS)

Nema, V. K.; Sharma, O. P.

1986-01-01

To improve upon the performance of chemical rockets, the nuclear reactor has been applied to a rocket propulsion system using hydrogen gas as working fluid and a graphite-composite forming a part of the structure. Under the boundary layer approximation, theoretical predictions of skin friction coefficient, surface heat transfer rate and surface regression rate have been made for laminar/turbulent dissociated hydrogen gas flowing over a flat graphite surface. The external stream is assumed to be frozen. The analysis is restricted to Mach numbers low enough to deal with the situation of only surface-reaction between hydrogen and graphite. Empirical correlations of displacement thickness, local skin friction coefficient, local Nusselt number and local non-dimensional heat transfer rate have been obtained. The magnitude of the surface regression rate is found low enough to ensure the use of graphite as a linear or a component of the system over an extended period without loss of performance.

Damage tolerance of nuclear graphite at elevated temperatures

DOE PAGES

Liu, Dong; Gludovatz, Bernd; Barnard, Harold S.; ...

2017-06-30

Nuclear-grade graphite is a critically important high-temperature structural material for current and potentially next generation of fission reactors worldwide. It is imperative to understand its damage-tolerant behaviour and to discern the mechanisms of damage evolution under in-service conditions. Here we perform in situ mechanical testing with synchrotron X-ray computed micro-tomography at temperatures between ambient and 1,000 °C on a nuclear-grade Gilsocarbon graphite. We find that both the strength and fracture toughness of this graphite are improved at elevated temperature. Whereas this behaviour is consistent with observations of the closure of microcracks formed parallel to the covalent-sp 2-bonded graphene layers atmore » higher temperatures, which accommodate the more than tenfold larger thermal expansion perpendicular to these layers, we attribute the elevation in strength and toughness primarily to changes in the residual stress state at 800–1,000 °C, specifically to the reduction in significant levels of residual tensile stresses in the graphite that are ‘frozen-in’ following processing.« less

Damage tolerance of nuclear graphite at elevated temperatures

PubMed Central

Liu, Dong; Gludovatz, Bernd; Barnard, Harold S.; Kuball, Martin; Ritchie, Robert O.

2017-01-01

Nuclear-grade graphite is a critically important high-temperature structural material for current and potentially next generation of fission reactors worldwide. It is imperative to understand its damage-tolerant behaviour and to discern the mechanisms of damage evolution under in-service conditions. Here we perform in situ mechanical testing with synchrotron X-ray computed micro-tomography at temperatures between ambient and 1,000 °C on a nuclear-grade Gilsocarbon graphite. We find that both the strength and fracture toughness of this graphite are improved at elevated temperature. Whereas this behaviour is consistent with observations of the closure of microcracks formed parallel to the covalent-sp2-bonded graphene layers at higher temperatures, which accommodate the more than tenfold larger thermal expansion perpendicular to these layers, we attribute the elevation in strength and toughness primarily to changes in the residual stress state at 800–1,000 °C, specifically to the reduction in significant levels of residual tensile stresses in the graphite that are ‘frozen-in’ following processing. PMID:28665405

Safe and Effective Deactivation of Metallic Sodium Filled Scrap and Cold Traps From Sodium-cooled Nuclear Reactor D and D - 12176

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

Nester, Dean; Crocker, Ben; Smart, Bill

2012-07-01

As part of the Plateau Remediation Project at US Department of Energy's Hanford, Washington site, CH2M Hill Plateau Remediation Company (CHPRC) contracted with IMPACT Services, LLC to receive and deactivate approximately 28 cubic meters of sodium metal contaminated debris from two sodium-cooled research reactors (Enrico Fermi Unit 1 and the Fast Flux Test Facility) which had been stored at Hanford for over 25 years. CHPRC found an off-site team composed of IMPACT Services and Commodore Advanced Sciences, Inc., with the facilities and technological capabilities to safely and effectively perform deactivation of this sodium metal contaminated debris. IMPACT Services provided themore » licensed fixed facility and the logistical support required to receive, store, and manage the waste materials before treatment, and the characterization, manifesting, and return shipping of the cleaned material after treatment. They also provided a recycle outlet for the liquid sodium hydroxide byproduct resulting from removal of the sodium from reactor parts. Commodore Advanced Sciences, Inc. mobilized their patented AMANDA unit to the IMPACT Services site and operated the unit to perform the sodium removal process. Approximately 816 Kg of metallic sodium were removed and converted to sodium hydroxide, and the project was accomplished in 107 days, from receipt of the first shipment at the IMPACT Services facility to the last outgoing shipment of deactivated scrap metal. There were no safety incidents of any kind during the performance of this project. The AMANDA process has been demonstrated in this project to be both safe and effective for deactivation of sodium and NaK. It has also been used in other venues to treat other highly reactive alkali metals, such as lithium (Li), potassium (K), NaK and Cesium (Cs). (authors)« less

ICP-MS measurement of silver diffusion coefficient in graphite IG-110 between 1048K and 1284K

NASA Astrophysics Data System (ADS)

Carter, L. M.; Seelig, J. D.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2018-01-01

Silver-110m has been shown to permeate intact silicon carbide and pyrolytic carbon coating layers of the TRISO fuel particles during normal High Temperature Gas-Cooled Reactor (HTGR) operational conditions. The diffusion coefficients for silver in graphite IG-110 measured using a release method designed to simulate HTGR conditions of high temperature and flowing helium in the temperature range 1048-1253 K are reported. The measurements were made using spheres milled from IG-110 graphite that were infused with silver using a pressure vessel technique. The Ag diffusion was measured using a time release technique with an ICP-MS instrument for detection. The results of this work are:

FY 2017-Influence of Sodium Environment on the Tensile Properties of Advanced Alloys

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

Natesan, K.; Li, Meimei; Chen, Wei-Ying

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602093), under the Work Package AT-17AN160209, “Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program. Three laboratory-size heats of Alloy 709 austenitic steel were investigated in liquid sodium environments at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. In addition, a commercial scale heat has beenmore » produced and hot-rolled into plates.« less

Monitoring of tritium

DOEpatents

Corbett, James A.; Meacham, Sterling A.

1981-01-01

The fluid from a breeder nuclear reactor, which may be the sodium cooling fluid or the helium reactor-cover-gas, or the helium coolant of a gas-cooled reactor passes over the portion of the enclosure of a gaseous discharge device which is permeable to hydrogen and its isotopes. The tritium diffused into the discharge device is radioactive producing beta rays which ionize the gas (argon) in the discharge device. The tritium is monitored by measuring the ionization current produced when the sodium phase and the gas phase of the hydrogen isotopes within the enclosure are in equilibrium.

A new oxidation based technique for artifact free TEM specimen preparation of nuclear graphite

NASA Astrophysics Data System (ADS)

Johns, Steve; Shin, Wontak; Kane, Joshua J.; Windes, William E.; Ubic, Rick; Karthik, Chinnathambi

2018-07-01

Graphite is a key component in designs of current and future nuclear reactors whose in-service lifetimes are dependent upon the mechanical performance of the graphite. Irradiation damage from fast neutrons creates lattice defects which have a dynamic effect on the microstructure and mechanical properties of graphite. Transmission electron microscopy (TEM) can offer real-time monitoring of the dynamic atomic-level response of graphite subjected to irradiation; however, conventional TEM specimen-preparation techniques, such as argon ion milling itself, damage the graphite specimen and introduce lattice defects. It is impossible to distinguish these defects from the ones created by electron or neutron irradiation. To ensure that TEM specimens are artifact-free, a new oxidation-based technique has been developed. Bulk nuclear grades of graphite (IG-110 and NBG-18) and highly oriented pyrolytic graphite (HOPG) were initially mechanically thinned to ∼60 μm. Discs 3 mm in diameter were then oxidized at temperatures between 575 °C and 625 °C in oxidizing gasses using a new jet-polisher-like set-up in order to achieve optimal oxidation conditions to create self-supporting electron-transparent TEM specimens. The quality of these oxidized specimens were established using optical and electron microscopy. Samples oxidized at 575 °C exhibited large areas of electron transparency and the corresponding lattice imaging showed no apparent damage to the graphite lattice.

A new oxidation based technique for artifact free TEM specimen preparation of nuclear graphite

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

Johns, Steve; Shin, Wontak; Kane, Joshua J.

Graphite is a key component in designs of current and future nuclear reactors whose in-service lifetimes are dependent upon the mechanical performance of the graphite. Irradiation damage from fast neutrons creates lattice defects which have a dynamic effect on the microstructure and mechanical properties of graphite. Transmission electron microscopy (TEM) can offer real-time monitoring of the dynamic atomic-level response of graphite subjected to irradiation; however, conventional TEM specimen-preparation techniques, such as argon ion milling itself, damage the graphite specimen and introduce lattice defects. It is impossible to distinguish these defects from the ones created by electron or neutron irradiation. Thus,tomore » ensure that TEM specimens are artifact-free, a new oxidation-based technique has been developed. Bulk nuclear grades of graphite (IG-110 and NBG-18) and highly oriented pyrolytic graphite (HOPG) were initially mechanically thinned to ~60μm. Discs 3mm in diameter were then oxidized at temperatures between 575°C and 625°C in oxidizing gasses using a new jet-polisher-like set-up in order to achieve optimal oxidation conditions to create self-supporting electron-transparent TEM specimens. The quality of these oxidized specimens were established using optical and electron microscopy. Samples oxidized at 575°C exhibited large areas of electron transparency and the corresponding lattice imaging showed no apparent damage to the graphite lattice.« less

A new oxidation based technique for artifact free TEM specimen preparation of nuclear graphite

DOE PAGES

Johns, Steve; Shin, Wontak; Kane, Joshua J.; ...

2018-04-03

Graphite is a key component in designs of current and future nuclear reactors whose in-service lifetimes are dependent upon the mechanical performance of the graphite. Irradiation damage from fast neutrons creates lattice defects which have a dynamic effect on the microstructure and mechanical properties of graphite. Transmission electron microscopy (TEM) can offer real-time monitoring of the dynamic atomic-level response of graphite subjected to irradiation; however, conventional TEM specimen-preparation techniques, such as argon ion milling itself, damage the graphite specimen and introduce lattice defects. It is impossible to distinguish these defects from the ones created by electron or neutron irradiation. Thus,tomore » ensure that TEM specimens are artifact-free, a new oxidation-based technique has been developed. Bulk nuclear grades of graphite (IG-110 and NBG-18) and highly oriented pyrolytic graphite (HOPG) were initially mechanically thinned to ~60μm. Discs 3mm in diameter were then oxidized at temperatures between 575°C and 625°C in oxidizing gasses using a new jet-polisher-like set-up in order to achieve optimal oxidation conditions to create self-supporting electron-transparent TEM specimens. The quality of these oxidized specimens were established using optical and electron microscopy. Samples oxidized at 575°C exhibited large areas of electron transparency and the corresponding lattice imaging showed no apparent damage to the graphite lattice.« less

Preliminary Design of Critical Function Monitoring System of PGSFR

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

NONE

2015-07-01

A PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) is under development at Korea Atomic Energy Research Institute. A critical function monitoring system of the PGSFR is preliminarily studied. The functions of CFMS are to display critical plant variables related to the safety of the plant during normal and accident conditions and guide the operators corrective actions to keep the plant in a safe condition and mitigate the consequences of accidents. The minimal critical functions of the PGSFR are composed of reactivity control, reactor core cooling, reactor coolant system integrity, primary heat transfer system(PHTS) heat removal, sodium water reaction mitigation, radiation controlmore » and containment conditions. The variables and alarm legs of each critical function of the PGSFR are as follows; - Reactivity control: The variables of reactivity control function are power range neutron flux instrumentation, intermediate range neutron flux instrumentation, source range neutron flux instrumentation, and control rod bottom contacts. The alarm leg to display the reactivity controls consists of status of control drop malfunction, high post trip power and thermal reactivity addition. - Reactor core cooling: The variables are PHTS sodium level, hot pool temperature of PHTS, subassembly exit temperature, cold pool temperature of the PHTS, PHTS pump current, and PHTS pump breaker status. The alarm leg consists of high core delta temperature, low sodium level of the PHTS, high subassembly exit temperature, and low PHTS pump load. - Reactor coolant system integrity: The variables are PHTS sodium level, cover gas pressure, and safeguard vessel sodium level. The alarm leg is composed of low sodium level of PHTS, high cover gas pressure and high sodium level of the safety guard vessel. - PHTS heat removal: The variables are PHTS sodium level, hot pool temperature of PHTS, core exit temperature, cold pool temperature of the PHTS, flow rate of passive residual heat removal system, flow rate of active residual heat removal system, and temperatures of air heat exchanger temperature of residual heat removal systems. The alarm legs are composed of two legs of a 'passive residual heat removal system not cooling' and 'active residual heat removal system not cooling'. - Sodium water reaction mitigation: The variables are intermediate heat transfer system(IHTS) pressure, pressure and temperature and level of sodium dump tank, the status of rupture disk, hydrogen concentration in IHTS and direct variable of sodium-water-reaction measure. The alarm leg consists of high IHTS pressure, the status of sodium water reaction mitigation system and the indication of direct measure. - Radiation control: The variables are radiation of PHTS, radiation of IHTS, and radiation of containment purge. The alarm leg is composed of high radiation of PHTS and IHTS, and containment purge system. - Containment condition: The variables are containment pressure, containment isolation status, and sodium fire. The alarm leg consists of high containment pressure, status of containment isolation and status of sodium fire. (authors)« less

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

Campbell, Anne A.; Katoh, Yutai; Snead, Mary A.

A new, fine-grain nuclear graphite, grade G347A from Tokai Carbon Co., Ltd., has been irradiated in the High Flux Isotope Reactor at Oak Ridge National Laboratory to study the materials property changes that occur when exposed to neutron irradiation at temperatures of interest for Generation-IV nuclear reactor applications. Specimen temperatures ranged from 290°C to 800 °C with a maximum neutron fluence of 40 × 10 25 n/m 2 [E > 0.1 MeV] (~30dpa). Lastly, observed behaviors include: anisotropic behavior of dimensional change in an isotropic graphite, Young's modulus showing parabolic fluence dependence, electrical resistivity increasing at low fluence and additionalmore » increase at high fluence, thermal conductivity rapidly decreasing at low fluence followed by continued degradation, and a similar plateau value of the mean coefficient of thermal expansion for all irradiation temperatures.« less

Recovery of Iron from Chromium Vanadium-Bearing Titanomagnetite Concentrate by Direct Reduction

NASA Astrophysics Data System (ADS)

Wang, Mingyu; Zhou, Shengfan; Wang, Xuewen; Chen, Bianfang; Yang, Haoxiang; Wang, Saikui; Luo, Pengfei

2016-10-01

The recovery of iron from chromium vanadium-bearing titanomagnetite concentrate was investigated by direct reduction, followed by magnetic separation. The results indicated that the metallization rate of iron can reach 98.9% at a temperature of 1200°C for a reduction duration of 60 min with the addition of 16% graphite powder and 0.5% sodium oxalate. Although the addition of borax, sodium carbonate and sodium oxalate to the chromium vanadium-bearing titanomagnetite concentrate can all improve the metallization rate of iron, the effect of sodium oxalate was the best. Sodium oxalate not only increases the metallization rate of iron but also promotes the growth of metallic iron. After magnetic separating, the recovery of iron was 92.8% and the iron content of magnetic concentrate was 88.4%.

Thermal neutron streaming effects and WIMS analysis of the Penn State subcritical graphite pile

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

Feltus, M.A.; Zediak, C.S.; Jester, W.A.

1997-12-01

This analysis was performed on the Pennsylvania State University (PSU) subcritical reactor to find more accurate values for such nuclear parameters as the thermal fuel utilization factor, thermal diffusion length in the graphite, migration area, k{sub eff}, etc. The analysis involved using the Winfrith Integrated Multigroup Scheme (WIMS) code as well as various hand calculations to find and compare those parameters. The data found in this analysis will be used by future students in the Penn State laboratory courses.

THE HOT CRITICAL ASSEMBLY $sub 4$CESAR$sub 4$ (in French)

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

Tanguy, P.

1963-07-01

With Cesar, the Cadarache Center for Nuclear Studies will be equipped with a zero-power critical assembly, which will enable it to obtain the data necessary for the development of natural uranium, graphite, gas reactors. Reactivity balance, evolution of the reactivity, and deformation of the flux curves are to be studied. These studies will complement those already being done on Marius, but carried out at room temperature; in Cesar the graphite temperature can reach 500 deg C. (auth)

SIKA—the multiplexing cold-neutron triple-axis spectrometer at ANSTO

NASA Astrophysics Data System (ADS)

Wu, C.-M.; Deng, G.; Gardner, J. S.; Vorderwisch, P.; Li, W.-H.; Yano, S.; Peng, J.-C.; Imamovic, E.

2016-10-01

SIKA is a new cold-neutron triple-axis spectrometer receiving neutrons from the cold source CG4 of the 20MW Open Pool Australian Light-water reactor. As a state-of-the-art triple-axis spectrometer, SIKA is equipped with a large double-focusing pyrolytic graphite monochromator, a multiblade pyrolytic graphite analyser and a multi-detector system. In this paper, we present the design, functions, and capabilities of SIKA, and discuss commissioning experimental results from powder and single-crystal samples to demonstrate its performance.

Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations

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

Vaughn, Kyle; Mausolff, Zander; Gonzalez, Esteban

Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.

Study of dynamics of glucose-glucose oxidase-ferricyanide reaction

NASA Astrophysics Data System (ADS)

Nováková, A.; Schreiberová, L.; Schreiber, I.

2011-12-01

This work is focused on dynamics of the glucose-glucose oxidase-ferricyanide enzymatic reaction with or without sodium hydroxide in a continuous-flow stirred tank reactor (CSTR) and in a batch reactor. This reaction exhibits pH-variations having autocatalytic character and is reported to provide nonlinear dynamic behavior (bistability, excitability). The dynamical behavior of the reaction was examined within a wide range of inlet parameters. The main inlet parameters were the ratio of concentrations of sodium hydroxide and ferricyanide and the flow rate. In a batch reactor we observed an autocatalytic drop of pH from slightly basic to medium acidic values. In a CSTR our aim was to find bistability in the presence of sodium hydroxide. However, only a basic steady state was found. In order to reach an acidic steady state, we investigated the system in the absence of sodium hydroxide. Under these conditions the transition from the basic to the acidic steady state was observed when inlet glucose concentration was increased.

Neutronics Analyses of the Minimum Original HEU TREAT Core

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

Kontogeorgakos, D.; Connaway, H.; Yesilyurt, G.

2014-04-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to validate the MCNP model of the TREAT reactor with the well-documented measurements which were taken during the start-up and early operation of TREAT. Furthermore, the effect of carbon graphitization was also addressed. The graphitization level was assumedmore » to be 100% (ANL/GTRI/TM-13/4). For this purpose, a set of experiments was chosen to validate the TREAT MCNP model, involving the approach to criticality procedure, in-core neutron flux measurements with foils, and isothermal temperature coefficient and temperature distribution measurements. The results of this study extended the knowledge base for the TREAT MCNP calculations and established the credibility of the MCNP model to be used in the core conversion feasibility analysis.« less

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

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

Palabrica, R.J.

1981-01-01

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

Cleanup Verification Package for the 118-F-1 Burial Ground

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

E. J. Farris and H. M. Sulloway

2008-01-10

This cleanup verification package documents completion of remedial action for the 118-F-1 Burial Ground on the Hanford Site. This burial ground is a combination of two locations formerly called Minor Construction Burial Ground No. 2 and Solid Waste Burial Ground No. 2. This waste site received radioactive equipment and other miscellaneous waste from 105-F Reactor operations, including dummy elements and irradiated process tubing; gun barrel tips, steel sleeves, and metal chips removed from the reactor; filter boxes containing reactor graphite chips; and miscellaneous construction solid waste.

A plasma arc reactor for fullerene research

NASA Astrophysics Data System (ADS)

Anderson, T. T.; Dyer, P. L.; Dykes, J. W.; Klavins, P.; Anderson, P. E.; Liu, J. Z.; Shelton, R. N.

1994-12-01

A modified Krätschmer-Huffman reactor for the mass production of fullerenes is presented. Fullerene mass production is fundamental for the synthesis of higher and endohedral fullerenes. The reactor employs mechanisms for continuous graphite-rod feeding and in situ slag removal. Soot collects into a Soxhlet extraction thimble which serves as a fore-line vacuum pump filter, thereby easing fullerene separation from soot. Thermal gravimetric analysis (TGA) for yield determination is reported. This TGA method is faster and uses smaller samples than Soxhlet extraction methods which rely on aromatic solvents. Production of 10 g of soot per hour is readily achieved utilizing this reactor. Fullerene yields of 20% are attained routinely.

Electroplating of the superconductive boride MgB2 from molten salts

NASA Astrophysics Data System (ADS)

Abe, Hideki; Yoshii, Kenji; Nishida, Kenji; Imai, Motoharu; Kitazawa, Hideaki

2005-02-01

An electroplating technique of the superconductive boride MgB2 onto graphite substrates is reported. Films of MgB2 with a thickness of tens micrometer were fabricated on the planar and curved surfaces of graphite substrates by means of electrolysis on a mixture of magnesium chloride, potassium chloride, sodium chloride, and magnesium borate fused at 600 °C under an Ar atmosphere. The electrical resistivity and magnetization measurements revealed that the electroplated MgB2 films undergo a superconducting transition with the critical temperature (Tc) of 36 K.

Energy Distributions of Neutrons Scattered from Graphite, Light and Heavy Water, Ice, Zirconium Hydride, Lithium Hydride, Sodium Hydride and Ammonium Chloride by the Beryllium Detector Method

DOE R&D Accomplishments Database

Woods, A. D. B.; Brockhouse, Bertram N.; Sakamoto, M.; Sinclair, R. N.

1960-09-12

Energy distributions of neutrons scattered from various moderators and from several hydrogenous substances were measured at energy transfers of 0.02 to 0.24 ev. Results from experiments on graphite, light and heavy water, ice, ZrH, LiH, NaH, and NH4Cl are included. It is noted that the results are of a preliminary character; however, they are probably the most accurate measurements of high-energy transfers yet made. (J.R.D.)

Experimental Study and Computational Simulations of Key Pebble Bed Thermo-mechanics Issues for Design and Safety

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

Tokuhiro, Akira; Potirniche, Gabriel; Cogliati, Joshua

2014-07-08

An experimental and computational study, consisting of modeling and simulation (M&S), of key thermal-mechanical issues affecting the design and safety of pebble-bed (PB) reactors was conducted. The objective was to broaden understanding and experimentally validate thermal-mechanic phenomena of nuclear grade graphite, specifically, spheres in frictional contact as anticipated in the bed under reactor relevant pressures and temperatures. The contact generates graphite dust particulates that can subsequently be transported into the flowing gaseous coolent. Under postulated depressurization transients and with the potential for leaked fission products to be adsorbed onto graphite 'dust', there is the potential for fission products to escapemore » from the primary volume. This is a design safety concern. Furthermore, earlier safety assessment identified the distinct possibility for the dispersed dust to combust in contact with air if sufficient conditions are met. Both of these phenomena were noted as important to design review and containing uncertainty to warrant study. The team designed and conducted two separate effects tests to study and benchmark the potential dust-generation rate, as well as study the conditions under which a dust explosion may occure in a standardized, instrumented explosion chamber.« less

Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

DOEpatents

Brehm, Jr., William F.; Colburn, Richard P.

1982-01-01

An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

Preliminary design of high temperature ultrasonic transducers for liquid sodium environments

NASA Astrophysics Data System (ADS)

Prowant, M. S.; Dib, G.; Qiao, H.; Good, M. S.; Larche, M. R.; Sexton, S. S.; Ramuhalli, P.

2018-04-01

Advanced reactor concepts include fast reactors (including sodium-cooled fast reactors), gas-cooled reactors, and molten-salt reactors. Common to these concepts is a higher operating temperature (when compared to light-water-cooled reactors), and the proposed use of new alloys with which there is limited operational experience. Concerns about new degradation mechanisms, such as high-temperature creep and creep fatigue, that are not encountered in the light-water fleet and longer operating cycles between refueling intervals indicate the need for condition monitoring technology. Specific needs in this context include periodic in-service inspection technology for the detection and sizing of cracking, as well as technologies for continuous monitoring of components using in situ probes. This paper will discuss research on the development and evaluation of high temperature (>550°C; >1022°F) ultrasonic probes that can be used for continuous monitoring of components. The focus of this work is on probes that are compatible with a liquid sodium-cooled reactor environment, where the core outlet temperatures can reach 550°C (1022°F). Modeling to assess sensitivity of various sensor configurations and experimental evaluation have pointed to a preferred design and concept of operations for these probes. This paper will describe these studies and ongoing work to fabricate and fully evaluate survivability and sensor performance over extended periods at operational temperatures.

Updating irradiated graphite disposal: Project 'GRAPA' and the international decommissioning network.

PubMed

Wickham, Anthony; Steinmetz, Hans-Jürgen; O'Sullivan, Patrick; Ojovan, Michael I

2017-05-01

Demonstrating competence in planning and executing the disposal of radioactive wastes is a key factor in the public perception of the nuclear power industry and must be demonstrated when making the case for new nuclear build. This work addresses the particular waste stream of irradiated graphite, mostly derived from reactor moderators and amounting to more than 250,000 tonnes world-wide. Use may be made of its unique chemical and physical properties to consider possible processing and disposal options outside the normal simple classifications and repository options for mixed low or intermediate-level wastes. The IAEA has an obvious involvement in radioactive waste disposal and has established a new project 'GRAPA' - Irradiated Graphite Processing Approaches - to encourage an international debate and collaborative work aimed at optimising and facilitating the treatment of irradiated graphite. Copyright © 2017 Elsevier Ltd. All rights reserved.

Irradiation Creep in Graphite

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

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13

An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarlymore » characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.« less

Neutron fluxes in test reactors

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

Youinou, Gilles Jean-Michel

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

Status of Initial Assessment of Physical and Mechanical Properties of Graphite Grades for NGNP Appkications

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

Strizak, Joe P; Burchell, Timothy D; Windes, Will

2011-12-01

Current candidate graphite grades for the core structures of NGNP include grades NBG-17, NBG-18, PCEA and IG-430. Both NBG-17 and NBG-18 are manufactured using pitch coke, and are vibrationally molded. These medium grain products are produced by SGL Carbon SAS (France). Tayo Tanso (Japan) produces IG-430 which is a petroleum coke, isostatically molded, nuclear grade graphite. And PCEA is a medium grain, extruded graphite produced by UCAR Carbon Co. (USA) from petroleum coke. An experimental program has been initiated to develop physical and mechanical properties data for these current candidate graphites. The results will be judged against the requirements formore » nuclear grade graphites set forth in ASTM standard D 7219-05 "Standard Specification for Isotropic and Near-isotropic Nuclear Graphites". Physical properties data including thermal conductivity and coefficient of thermal expansion, and mechanical properties data including tensile, compressive and flexural strengths will be obtained using the established test methods covered in D-7219 and ASTM C 781-02 "Standard Practice for Testing Graphite and Boronated Graphite Components for High-Temperature Gas-Cooled Nuclear Reactors". Various factors known to effect the properties of graphites will be investigated. These include specimen size, spatial location within a graphite billet, specimen orientation (ag and wg) within a billet, and billet-to-billet variations. The current status of the materials characterization program is reported herein. To date billets of the four graphite grades have been procured, and detailed cut up plans for obtaining the various specimens have been prepared. Particular attention has been given to the traceability of each specimen to its spatial location and orientation within a billet.« less

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

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

Bess, John D.

2014-03-01

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

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

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

John D. Bess

2013-03-01

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

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

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

John D. Bess

2013-03-01

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

DE-NE0008277_PROTEUS final technical report 2018

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

Enqvist, Andreas

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

Investigation of Isotopically Tailored Boron in Advanced Fission and Fusion Reactor Systems.

NASA Astrophysics Data System (ADS)

Domaszek, Gerald Raymond

This research examines the use of B^ {11}, in the form of metallic boron and boron carbide, as a moderating and reflecting material. An examination of the neutronic characteristics of the B ^{11} isotope of boron has revealed that B^{11} has neutron scattering and absorption cross sections favorably comparable to those of Be^9 and C^ {12}. Preliminary analysis of the neutronics of B ^{11} were performed by conducting one dimensional transport calculations on an infinite slab of varying thickness. Beryllium is the best of the three materials in reflecting neutrons due primarily to the contribution from (n,2n) reactions. Tailored neutron energy beam transmission experiments were carried out to experimentally verify the predicted neutronic characteristics of B^{11 }. To further examine the neutron moderating and reflecting characteristics of B^{11 }, the energy dependent neutron flux was measured as a function of position in an exponential pile constructed of B_4C isotopically enriched to 98.5 percent B^{11}. After the experimental verification of the neutronic behavior of B^{11}, further design studies were conducted using metallic boron and boron carbide enriched in the B^{11 } isotope. The use of materials isotopically enriched in B^{11} as a liner in the first wall/blanket of a magnetic confinement fusion reactor demonstrated acceptable tritium regeneration in the lithium blanket. Analysis of the effect of contaminant levels of B^{10} showed that B^{10} contents of less than 1 percent in metallic boron produced negligible adverse effects on the tritium breeding. A comparison of the effectiveness of graphite and B^{11}_4C when used as moderators in a reactor fueled with natural uranium has shown that the maximum k_infty for a given fuel rod design is approximately the same for both materials. Approximately half the volume of the moderator is required when B^{11 }_4C is substituted for graphite to obtain essentially the same K_infty . An analysis of the effectiveness of various materials as reflector control elements for a compact space reactor has shown that B^{11} is neutronically superior to graphite in these applications. Metallic boron and boron carbide isotopically enriched in B^{11} have been demonstrated to be neutronically acceptable for varied applications in advanced reactor systems. B^ {11} has been shown to be superior in performance to graphite. While only somewhat inferior to beryllium as neutron multipliers, B^ {11} and B^{11} _4C have safety, supply and cost advantage over beryllium. (Abstract shortened with permission of author.).

Trace analysis of high-purity graphite by LA-ICP-MS.

PubMed

Pickhardt, C; Becker, J S

2001-07-01

Laser-ablation inductively coupled plasma mass spectrometry (LA-ICP-MS) has been established as a very efficient and sensitive technique for the direct analysis of solids. In this work the capability of LA-ICP-MS was investigated for determination of trace elements in high-purity graphite. Synthetic laboratory standards with a graphite matrix were prepared for the purpose of quantifying the analytical results. Doped trace elements, concentration 0.5 microg g(-1), in a laboratory standard were determined with an accuracy of 1% to +/- 7% and a relative standard deviation (RSD) of 2-13%. Solution-based calibration was also used for quantitative analysis of high-purity graphite. It was found that such calibration led to analytical results for trace-element determination in graphite with accuracy similar to that obtained by use of synthetic laboratory standards for quantification of analytical results. Results from quantitative determination of trace impurities in a real reactor-graphite sample, using both quantification approaches, were in good agreement. Detection limits for all elements of interest were determined in the low ng g(-1) concentration range. Improvement of detection limits by a factor of 10 was achieved for analyses of high-purity graphite with LA-ICP-MS under wet plasma conditions, because the lower background signal and increased element sensitivity.

Fission Product Sorptivity in Graphite

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

Tompson, Jr., Robert V.; Loyalka, Sudarshan; Ghosh, Tushar

Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodatemore » the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one graduate student meant that data acquisition with the packed bed systems ended up competing for the graduate student’s available time with the electrodynamic balance redesign and assembly portions of the project. This competition for available time was eventually mitigated to some extent by the later recruitment of an undergraduate student to help with data collection using the packed bed system. It was only the recruitment of the second student that allowed the single particle balance design and construction efforts to proceed as far as they did during the project period. It should be added that some significant time was also spent by the graduate student cataloging previous work involving graphite. This eventually resulted in a review paper being submitted and accepted (“Adsorption of Iodine on Graphite in High Temperature Gas-Cooled Reactor Systems: A Review,” Kyle L. Walton, Tushar K. Ghosh, Dabir S. Viswanath, Sudarshan K. Loyalka, Robert V. Tompson). Our specific revised objectives in this project were as follows: Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using an EDB and a temperature controlled EDB; Experimentally obtain isotherms of Iodine for reactor grade IG-110 samples of graphite particles over a range of temperatures and pressures using a packed column bed apparatus; Explore the effect that charge has on the adsorption isotherms of iodine by varying the charges on and the voltages used to suspend the microscopic particles in the EDB; and To interpret these results in terms of the existing models (Langmuir, BET, Freundlich, and others) which we will modify as necessary to include charge related effects.« less

Fabrication methods and anisotropic properties of graphite matrix compacts for use in HTGR

NASA Astrophysics Data System (ADS)

Yeo, Sunghwan; Yun, Jihae; Kim, Sungok; Cho, Moon Sung; Lee, Young-Woo

2018-02-01

This study investigated the anisotropic microstructural, mechanical, and thermal properties of fabricated graphite matrix prismatic compacts for High Temperature Gas Cooled Reactor (HTGR) fuel. When the observed alignment of graphite grains and the coke derived from phenolic resin is in the transverse direction, the result is severely anisotropic thermal properties. Compacts with such orientation in the transverse direction exhibited increases of thermal expansion and conductivity up to 5.8 times and 4.82 times, respectively, more than those in the axial direction. The formation of pores due to the pyrolysis of phenolic resin was observed predominantly on upper region of the fabricated compacts. This anisotropic pore formation created anisotropic Vickers hardness on the planes with different directions.

METAL SHEATHED BODIES

DOEpatents

Finniston, H.M.; Wyatt, L.M.; Plail, O.S.

1961-06-27

An aluminum-cased uranium fuel element is patented for use in nuclear reactors. A layer of a substance such as graphite or a metallic film, preferably of relatively low thermal-neutron capture cross section, between the uranium and aluminum prevents their interdiffusion.

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

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

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

1978-06-01

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

The pre-conceptual design of the nuclear island of ASTRID

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

Saez, M.; Menou, S.; Uzu, B.

The CEA is involved in a substantial effort on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) pre-conceptual design in cooperation with EDF, as experienced Sodium-cooled Fast Reactor (SFR) operator, AREVA, as experienced SFR Nuclear Island engineering company and components designer, ALSTOM POWER as energy conversion system designer and COMEX NUCLEAIRE as mechanical systems designer. The CEA is looking for other partnerships, in France and abroad. The ASTRID preliminary design is based on a sodium-cooled pool reactor of 1500 MWth generating about 600 MWe, which is required to guarantee the representativeness of the reactor core and the main componentsmore » with regard to future commercial reactors. ASTRID lifetime target is 60 years. Two Energy Conversion Systems are studied in parallel until the end of 2012: Rankine steam cycle or Brayton gas based energy conversion cycle. ASTRID design is guided by the following major objectives: improved safety, simplification of structures, improved In Service Inspection and Repair (ISIR), improved manufacturing conditions for cost reduction and increased quality, reduction of risks related to sodium fires and water/sodium reaction, and improved robustness against external hazards. The core is supported by a diagrid, which lay on a strong back to transfer the weight to the main vessel. AREVA is involved in a substantial effort in order to improve the core support structure in particular regarding the ISIR and the connection to primary pump. In the preliminary design, the primary system is formed by the main vessel and the upper closure comprising the reactor roof, two rotating plugs - used for fuel handling - and the components plugs located in the roof penetrations. The Above Core Structure deflects the sodium flow in the hot pool and provides support to core instrumentation and guidance of the control rod drive mechanisms. The number of the major components in the main vessel, primary pumps, Intermediate Heat Exchangers, and Decay Heat Exchangers are now under consideration. Under normal conditions, power release is achieved using the steam/water plant (in case of Rankine steam cycle) or the gas plant (in case of Brayton gas cycle). The diverse design and operating modes of Decay Heat Removal systems provide protection against common cause failures. A Decay Heat Removal system through the reactor vault is in particular studied with the objective to complement Direct Reactor Cooling systems. At this stage of the studies, the secondary system comprises four independent sodium loops (two and three sodium loops configurations are also investigated). Each loop includes one mechanical pump (or a large capacity Annular Linear Induction Electromagnetic Pump), and three modular Steam Generator Units characterized by once through straight tube units with a ferritic tube bundle; nevertheless, helical coil steam generator with tubes made of Alloy 800, and inverted type steam generator with a ferritic tube bundle are also investigated. The limited power of each modular Steam Generator Unit allows the whole secondary loop to withstand a large water/sodium reaction consecutive to the postulated simultaneous rupture of all the heat exchange tubes of one module. The arrangement of the components is based on the 'Regain' concept, in which the secondary pump is situated at a low level in the circuit; conventional arrangement, as SUPERPHENIX type, is a back-up option. Alternative arrangements based on gas cycles are also studied together with Na-gas heat exchanger design. This paper presents a status of the ASTRID pre-conceptual design. The most promising options are highlighted as well as less risky and back-up options. (authors)« less

Study on preparation of SnO2-TiO2/Nano-graphite composite anode and electro-catalytic degradation of ceftriaxone sodium.

PubMed

Guo, Xiaolei; Wan, Jiafeng; Yu, Xiujuan; Lin, Yuhui

2016-12-01

In order to improve the electro-catalytic activity and catalytic reaction rate of graphite-like material, Tin dioxide-Titanium dioxide/Nano-graphite (SnO 2 -TiO 2 /Nano-G) composite was synthesized by a sol-gel method and SnO 2 -TiO 2 /Nano-G electrode was prepared in hot-press approach. The composite was characterized by X-ray photoelectron spectroscopy, fourier transform infrared, Raman, N 2 adsorption-desorption, scanning electrons microscopy, transmission electron microscopy and X-ray diffraction. The electrochemical performance of the SnO 2 -TiO 2 /Nano-G anode electrode was investigated via cyclic voltammetry and electrochemical impedance spectroscopy. The electro-catalytic performance was evaluated by the degradation of ceftriaxone sodium and the yield of ·OH radicals in the reaction system. The results demonstrated that TiO 2 , SnO 2 and Nano-G were composited successfully, and TiO 2 and SnO 2 particles dispersed on the surface and interlamination of the Nano-G uniformly. The specific surface area of SnO 2 modified anode was higher than that of TiO 2 /Nano-G anode and the degradation rate of ceftriaxone sodium within 120 min on SnO 2 -TiO 2 /Nano-G electrode was 98.7% at applied bias of 2.0 V. The highly efficient electro-chemical property of SnO 2 -TiO 2 /Nano-G electrode was attributed to the admirable conductive property of the Nano-G and SnO 2 -TiO 2 /Nano-G electrode. Moreover, the contribution of reactive species ·OH was detected, indicating the considerable electro-catalytic activity of SnO 2 -TiO 2 /Nano-G electrode. Copyright © 2016 Elsevier Ltd. All rights reserved.

Scoping studies of vapor behavior during a severe accident in a metal-fueling reactor

NASA Astrophysics Data System (ADS)

Spencer, B. W.; Marchaterre, J. F.

1985-04-01

The consequences of fuel melting and pin failures for a reactivity-insertion type accident in a sodium-cooled, pool-type reactor fueled with a metal alloy fuel were examined. The principal gas and vapor species released are shown to be Xe, Cs, and bond sodium contained within the fuel porosity. Condensation of sodium vapor as it expands into the upper sodium pool in a jet mixing regime may occur as rapidly as the vapor emerges from the disrupted core. If the predictions of rapid direct-contact condensation can be verified experimentally for the sodium system, the ability of vapor expansion to perform appreciable work on the system and the ability of an expanding vapor bubble to transport fuel and fission produce species to the cover gas region where they may be released to the containment are largely eliminated. The radionuclide species except for fission gas are largely retained within the core and sodium pool.

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

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

Ambrose, T.W.

1965-06-04

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

A Comparison of the Irradiation Creep Behavior of Several Graphites

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

Burchell, Timothy D; Windes, Will

2016-01-01

Graphite creep strain data from the irradiation creep capsule Advanced Graphite Creep-1 (AGC-1) are reported. This capsule was the first (prototype) of a series of five or six capsules planned as part of the AGC experiment, which was designed to fully characterize the effects of neutron irradiation and the radiation creep behavior of current nuclear graphite. The creep strain data and analysis are reported for the six graphite grades incorporated in the capsule. The AGC-1 capsule was irradiated in the Advanced Test Reactor at Idaho National Laboratory (INL) at approximately 700 C and to a peak dose of 7 dpamore » (displacements per atom). The specimen s final dose, temperature, and stress conditions have been reported by INL and were used during this analysis. The derived creep coefficients (K) were calculated for each grade and were found to compare well to literature data for the creep coefficient, even under the wide range of AGC-1 specimen temperatures. Comparisons were made between AGC-1 data and historical grade data for creep coefficients.« less

Alternate electrode materials for the SP100 reactor

NASA Astrophysics Data System (ADS)

Randich, E.

1992-05-01

This work was performed in response to a request by the Astro-Space Division of the General Electric Co. to develop alternate electrodes materials for the electrodes of the PD2 modules to be used in the SP100 thermoelectric power conversion system. Initially, the project consisted of four tasks: (1) development of a ZrB2 (C) CVD coating on SiMo substrates; (2) development of a ZrB2 (C) CVD coating on SiGe substrates; (3) development of CVI W for porous graphite electrodes; and (4) technology transfer of pertinent developed processes. The project evolved initially into developing only ZrB2 coatings on SiGe and graphite substrates, and later into developing ZrB2 coatings only on graphite substrates. Several sizes of graphite and pyrolytic carbon-coated graphite substrates were coated with ZrB2 during the project. For budgetary reasons, the project was terminated after half the allotted time had passed. Apart from the production of coated specimens for evaluation, the major accomplishment of the project was the development of the CVD processing to produce the desired coatings.

NEUTRONIC REACTOR

DOEpatents

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

1958-04-15

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

NEUTRONIC REACTOR SHIELD AND SPACER CONSTRUCTION

DOEpatents

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

1958-11-18

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

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

Snepvangers, J.J.M.

Equipment and results are described connected with irradiation studies of UO/sub 2/ fuels, fuel element testing in pressurized water loops, graphite irradiation, and steel irradiations with and without temperature control. The apparatus described is associated with a 20-Mw pool-type research reactor. (T.F.H.)

ETR BUILDING, TRA642, INTERIOR. BASEMENT. CUBICLE SHOWN IN ID33G101, ANOTHER ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. CUBICLE SHOWN IN ID-33-G-101, ANOTHER VIEW. PERSONNEL DOORWAY INTO CHAMBER IDENTIFIES SODIUM HAZARD AND POSSIBILITY OF INERT GAS. LIQUID SODIUM COOLANT WAS USED IN A SPECIAL ETR LOOP ADAPTED FOR IT IN 1972. INL NEGATIVE NO. HD24-3-2. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Theoretical Considerations for Improving the Pulse Power of a Battery through the Addition of a Second Electrochemically Active Material

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

Knehr, K. W.; West, Alan C.

Here, porous electrode theory is used to conduct case studies for when the addition of a second electrochemically active material can improve the pulse-power performance of an electrode. Case studies are conducted for the positive electrode of a sodium metal-halide battery and the graphite negative electrode of a lithium “rocking chair” battery. The replacement of a fraction of the nickel chloride capacity with iron chloride in a sodium metal-halide electrode and the replacement of a fraction of the graphite capacity with carbon black in a lithium-ion negative electrode were both predicted to increase the maximum pulse power by up tomore » 40%. In general, whether or not a second electrochemically active material increases the pulse power depends on the relative importance of ohmic-to-charge transfer resistances within the porous structure, the capacity fraction of the second electrochemically active material, and the kinetic and thermodynamic parameters of the two active materials.« less

Theoretical Considerations for Improving the Pulse Power of a Battery through the Addition of a Second Electrochemically Active Material

DOE PAGES

Knehr, K. W.; West, Alan C.

2016-05-26

Here, porous electrode theory is used to conduct case studies for when the addition of a second electrochemically active material can improve the pulse-power performance of an electrode. Case studies are conducted for the positive electrode of a sodium metal-halide battery and the graphite negative electrode of a lithium “rocking chair” battery. The replacement of a fraction of the nickel chloride capacity with iron chloride in a sodium metal-halide electrode and the replacement of a fraction of the graphite capacity with carbon black in a lithium-ion negative electrode were both predicted to increase the maximum pulse power by up tomore » 40%. In general, whether or not a second electrochemically active material increases the pulse power depends on the relative importance of ohmic-to-charge transfer resistances within the porous structure, the capacity fraction of the second electrochemically active material, and the kinetic and thermodynamic parameters of the two active materials.« less

Method of and apparatus for removing silicon from a high temperature sodium coolant

DOEpatents

Yunker, W.H.; Christiansen, D.W.

1983-11-25

This patent discloses a method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

Method of and apparatus for removing silicon from a high temperature sodium coolant

DOEpatents

Yunker, Wayne H.; Christiansen, David W.

1987-05-05

A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

Method of and apparatus for removing silicon from a high temperature sodium coolant

DOEpatents

Yunker, Wayne H.; Christiansen, David W.

1987-01-01

A method of and system for removing silicon from a high temperature liquid sodium coolant system for a nuclear reactor. The sodium is cooled to a temperature below the silicon saturation temperature and retained at such reduced temperature while inducing high turbulence into the sodium flow for promoting precipitation of silicon compounds and ultimate separation of silicon compound particles from the liquid sodium.

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1958-04-22

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

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

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

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

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

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

NASA Astrophysics Data System (ADS)

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

2009-08-01

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

In-Situ-Activated N-Doped Mesoporous Carbon from a Protic Salt and Its Performance in Supercapacitors.

PubMed

Mendes, Tiago C; Xiao, Changlong; Zhou, Fengling; Li, Haitao; Knowles, Gregory P; Hilder, Matthias; Somers, Anthony; Howlett, Patrick C; MacFarlane, Douglas R

2016-12-28

Protic salts have been recently recognized to be an excellent carbon source to obtain highly ordered N-doped carbon without the need of tedious and time-consuming preparation steps that are usually involved in traditional polymer-based precursors. Herein, we report a direct co-pyrolysis of an easily synthesized protic salt (benzimidazolium triflate) with calcium and sodium citrate at 850 °C to obtain N-doped mesoporous carbons from a single calcination procedure. It was found that sodium citrate plays a role in the final carbon porosity and acts as an in situ activator. This results in a large surface area as high as 1738 m 2 /g with a homogeneous pore size distribution and a moderate nitrogen doping level of 3.1%. X-ray photoelectron spectroscopy (XPS) measurements revealed that graphitic and pyridinic groups are the main nitrogen species present in the material, and their content depends on the amount of sodium citrate used during pyrolysis. Transmission electron microscopy (TEM) investigation showed that sodium citrate assists the formation of graphitic domains and many carbon nanosheets were observed. When applied as supercapacitor electrodes, a specific capacitance of 111 F/g in organic electrolyte was obtained and an excellent capacitance retention of 85.9% was observed at a current density of 10 A/g. At an operating voltage of 3.0 V, the device provided a maximum energy density of 35 W h/kg and a maximum power density of 12 kW/kg.

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

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

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

2011-07-01

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

Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

NASA Astrophysics Data System (ADS)

Blanchet, David; Fontaine, Bruno

2017-09-01

The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

Strange bedfellows: The curious case of STAR and Moata

NASA Astrophysics Data System (ADS)

Smith, A. M.; Levchenko, V. A.; Malone, G.

2013-01-01

The 2 MV tandem accelerator named ‘STAR’ was installed at ANSTO in 2003 and commissioned in 2004. It is used for ion beam analysis (IBA) and for radiocarbon measurements by accelerator mass spectrometry (AMS). Convenient space for the accelerator was found in the same building occupied by the decommissioned Argonaut-class nuclear reactor ‘Moata’; the name derives from the aboriginal word for ‘fire stick’ or ‘gentle fire’, appropriate for a 100 kW research reactor. This reactor operated between 1961 and 1995. In 2007 ANSTO’s Engineering Division assembled a team to dismantle and remove the reactor structure, along with its 12.1 tonnes of graphite reflector. The removal and remediation was completed in November 2010 and has won the team a number of prestigious awards. The entire operation was conducted inside a negatively-pressurised double-walled vinyl tent. An air curtain was positioned around the reactor core. The exhaust air from the tent passed through 2-stage HEPA filters before venting through an external stack. Neither ANSTO staff nor contractors received any significant radiation dose during the operation. Given the sensitivity of STAR for detection of 14C/12C (∼10-16) and the numerous routes for production of 14C in the reactor such as 13C(n, γ)14C, 14N(n, p)14C and 17O(n, α)14C there was the potential to directly contaminate the STAR environment with 14C. Furthermore, there was concern that reactor-14C could find its way from this building into the building where the radiocarbon sample preparation laboratories are located. This necessitated restrictions on staff movement between the buildings. We report on 14C control measurements made during and after the operation. These involved direct measurements on the reactor graphite and concrete bioshield, blank targets that were exposed in the building, swipe samples taken inside the tent and around the building and aerosol samples that were collected inside the building throughout the operation.

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation

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

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

2016-11-01

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation andmore » evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.« less

Camphor-mediated synthesis of carbon nanoparticles, graphitic shell encapsulated carbon nanocubes and carbon dots for bioimaging

PubMed Central

Oza, Goldie; Ravichandran, M.; Merupo, Victor-Ishrayelu; Shinde, Sachin; Mewada, Ashmi; Ramirez, Jose Tapia; Velumani, S.; Sharon, Madhuri; Sharon, Maheshwar

2016-01-01

A green method for an efficient synthesis of water-soluble carbon nanoparticles (CNPs), graphitic shell encapsulated carbon nanocubes (CNCs), Carbon dots (CDs) using Camphor (Cinnamomum camphora) is demonstrated. Here, we describe a competent molecular fusion and fission route for step-wise synthesis of CDs. Camphor on acidification and carbonization forms CNPs, which on alkaline hydrolysis form CNCs that are encapsulated by thick graphitic layers and on further reduction by sodium borohydride yielded CDs. Though excitation wavelength dependent photoluminescence is observed in all the three carbon nanostructures, CDs possess enhanced photoluminescent properties due to more defective carbonaceous structures. The surface hydroxyl and carboxyl functional groups make them water soluble in nature. They possess excellent photostability, higher quantum yield, increased absorption, decreased cytotoxicity and hence can be utilized as a proficient bio imaging agent. PMID:26905737

Camphor-mediated synthesis of carbon nanoparticles, graphitic shell encapsulated carbon nanocubes and carbon dots for bioimaging

NASA Astrophysics Data System (ADS)

Oza, Goldie; Ravichandran, M.; Merupo, Victor-Ishrayelu; Shinde, Sachin; Mewada, Ashmi; Ramirez, Jose Tapia; Velumani, S.; Sharon, Madhuri; Sharon, Maheshwar

2016-02-01

A green method for an efficient synthesis of water-soluble carbon nanoparticles (CNPs), graphitic shell encapsulated carbon nanocubes (CNCs), Carbon dots (CDs) using Camphor (Cinnamomum camphora) is demonstrated. Here, we describe a competent molecular fusion and fission route for step-wise synthesis of CDs. Camphor on acidification and carbonization forms CNPs, which on alkaline hydrolysis form CNCs that are encapsulated by thick graphitic layers and on further reduction by sodium borohydride yielded CDs. Though excitation wavelength dependent photoluminescence is observed in all the three carbon nanostructures, CDs possess enhanced photoluminescent properties due to more defective carbonaceous structures. The surface hydroxyl and carboxyl functional groups make them water soluble in nature. They possess excellent photostability, higher quantum yield, increased absorption, decreased cytotoxicity and hence can be utilized as a proficient bio imaging agent.

Electrochemical lithiation performance and characterization of silicon-graphite composites with lithium, sodium, potassium, and ammonium polyacrylate binders.

PubMed

Han, Zhen-Ji; Yamagiwa, Kiyofumi; Yabuuchi, Naoaki; Son, Jin-Young; Cui, Yi-Tao; Oji, Hiroshi; Kogure, Akinori; Harada, Takahiro; Ishikawa, Sumihisa; Aoki, Yasuhito; Komaba, Shinichi

2015-02-07

Poly(acrylic acid) (PAH), which is a water soluble polycarboxylic acid, is neutralized by adding different amounts of LiOH, NaOH, KOH, and ammonia (NH4OH) aqueous solutions to fix neutralization degrees. The differently neutralized polyacid, alkali and ammonium polyacrylates are examined as polymeric binders for the preparation of Si-graphite composite electrodes as negative electrodes for Li-ion batteries. The electrode performance of the Si-graphite composite depends on the alkali chemicals and neutralization degree. It is found that 80% NaOH-neutralized polyacrylate binder (a pH value of the resultant aqueous solution is ca. 6.7) is the most efficient binder to enhance the electrochemical lithiation and de-lithiation performance of the Si-graphite composite electrode compared to that of conventional PVdF and the other binders used in this study. The optimum polyacrylate binder highly improves the dispersion of active material in the composite electrode. The binder also provides the strong adhesion, suitable porosity, and hardness for the composite electrode with 10% (m/m) binder content, resulting in better electrochemical reversibility. From these results, the factors of alkali-neutralized polyacrylate binders affecting the electrode performance of Si-graphite composite electrodes are discussed.

Characterization of nuclear graphite elastic properties using laser ultrasonic methods

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

Zeng, Fan W; Han, Karen; Olasov, Lauren R

2015-01-01

Laser ultrasonic methods have been used to characterize the elastic behaviors of commercially-available and legacy nuclear graphites. Since ultrasonic techniques are sensitive to various aspects of graphite microstructure including preferred grain orientation, microcrack orientation and porosity, laser ultrasonics is a candidate technique for monitoring graphite degradation and structural integrity in environments expected in high-temperature, gas-cooled nuclear reactors. Aspects of materials texture can be assessed by studying ultrasonic wavespeeds as a function of propagation direction and polarization. Shear wave birefringence measurements, in particular, can be used to evaluate elastic anisotropy. In this work, laser ultrasonic measurements of graphite moduli have beenmore » made to provide insight into the relationship between the microstructures and the macroscopic stiffnesses of these materials. In particular, laser ultrasonic measurements have been made using laser line sources to produce shear waves with specific polarizations. By varying the line orientation relative to the sample, shear wave birefringence measurements have been recorded. Results from shear wave birefringence measurements show that an isostatically molded graphite, such as PCIB, behaves isotropically, while an extruded graphite, such as H-451, displays significant ultrasonic texture. Graphites have complicated microstructures that depend on the manufacturing processes used, and ultrasonic texture in these materials could originate from grain orientation and preferred microcrack alignment. Effects on material isotropy due to service related microstructural changes are possible and the ultimate aim of this work is to determine the degree to which these changes can be assessed nondestructively using laser ultrasonics measurements« less

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

Li, Meimei; Natesan, K.; Chen, Wei-Ying

This report provides an update on the understanding of the effect of sodium exposures on microstructure and tensile properties of Grade 91 (G91) steel in support of the design and operation of G91 components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602018), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

Corrosion of 316 stainless steel in high temperature molten Li2BeF4 (FLiBe) salt

NASA Astrophysics Data System (ADS)

Zheng, Guiqiu; Kelleher, Brian; Cao, Guoping; Anderson, Mark; Allen, Todd; Sridharan, Kumar

2015-06-01

In support of structural material development for the fluoride-salt-cooled high-temperature reactor (FHR), corrosion tests of 316 stainless steel were performed in the potential primary coolant, molten Li2BeF4 (FLiBe) at 700 °C for an exposure duration up to 3000 h. Tests were performed in both 316 stainless steel and graphite capsules. Corrosion in both capsule materials occurred by the dissolution of chromium from the stainless steel into the salt which led to the depletion of chromium predominantly along the grain boundaries of the test samples. The samples tested in graphite capsules showed a factor of two greater depth of corrosion attack as measured in terms of chromium depletion, compared to those tested in 316 stainless steel capsules. The samples tested in graphite capsules showed the formation of Cr7C3 particulate phases throughout the depth of the corrosion layer. Samples tested in both types of capsule materials showed the formation of MoSi2 phase due to increased activity of Mo and Si as a result of Cr depletion, and furthermore corrosion promoted the formation of a α-ferrite phase in the near-surface regions of the 316 stainless steel. Based on the corrosion tests, the corrosion attack depth in FLiBe salt was predicted as 17.1 μm/year and 31.2 μm/year for 316 stainless steel tested in 316 stainless steel and in graphite capsules respectively. It is in an acceptable range compared to the Hastelloy-N corrosion in the Molten Salt Reactor Experiment (MSRE) fuel salt.

Effects of Oxidation on Oxidation-Resistant Graphite

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

Windes, William; Smith, Rebecca; Carroll, Mark

2015-05-01

The Advanced Reactor Technology (ART) Graphite Research and Development Program is investigating doped nuclear graphite grades that exhibit oxidation resistance through the formation of protective oxides on the surface of the graphite material. In the unlikely event of an oxygen ingress accident, graphite components within the VHTR core region are anticipated to oxidize so long as the oxygen continues to enter the hot core region and the core temperatures remain above 400°C. For the most serious air-ingress accident which persists over several hours or days the continued oxidation can result in significant structural damage to the core. Reducing the oxidationmore » rate of the graphite core material during any air-ingress accident would mitigate the structural effects and keep the core intact. Previous air oxidation testing of nuclear-grade graphite doped with varying levels of boron-carbide (B4C) at a nominal 739°C was conducted for a limited number of doped specimens demonstrating a dramatic reduction in oxidation rate for the boronated graphite grade. This report summarizes the conclusions from this small scoping study by determining the effects of oxidation on the mechanical strength resulting from oxidation of boronated and unboronated graphite to a 10% mass loss level. While the B4C additive did reduce mechanical strength loss during oxidation, adding B4C dopants to a level of 3.5% or more reduced the as-fabricated compressive strength nearly 50%. This effectively minimized any benefits realized from the protective film formed on the boronated grades. Future work to infuse different graphite grades with silicon- and boron-doped material as a post-machining conditioning step for nuclear components is discussed as a potential solution for these challenges in this report.« less

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

NASA Astrophysics Data System (ADS)

Guidez, Joel; Saturnin, Anne

2017-11-01

During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

Surface Anchoring of Nematic Phase on Carbon Nanotubes: Nanostructure of Ultra-High Temperature Materials

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

Ogale, Amod A

2012-04-27

Nuclear energy is a dependable and economical source of electricity. Because fuel supply sources are available domestically, nuclear energy can be a strong domestic industry that can reduce dependence on foreign energy sources. Commercial nuclear power plants have extensive security measures to protect the facility from intruders [1]. However, additional research efforts are needed to increase the inherent process safety of nuclear energy plants to protect the public in the event of a reactor malfunction. The next generation nuclear plant (NGNP) is envisioned to utilize a very high temperature reactor (VHTR) design with an operating temperature of 650-1000°C [2]. Onemore » of the most important safety design requirements for this reactor is that it must be inherently safe, i.e., the reactor must shut down safely in the event that the coolant flow is interrupted [2]. This next-generation Gen IV reactor must operate in an inherently safe mode where the off-normal temperatures may reach 1500°C due to coolant-flow interruption. Metallic alloys used currently in reactor internals will melt at such temperatures. Structural materials that will not melt at such ultra-high temperatures are carbon/graphtic fibers and carbon-matrix composites. Graphite does not have a measurable melting point; it is known to sublime starting about 3300°C. However, neutron radiation-damage effects on carbon fibers are poorly understood. Therefore, the goal of this project is to obtain a fundamental understanding of the role of nanotexture on the properties of resulting carbon fibers and their neutron-damage characteristics. Although polygranular graphite has been used in nuclear environment for almost fifty years, it is not suitable for structural applications because it do not possess adequate strength, stiffness, or toughness that is required of structural components such as reaction control-rods, upper plenum shroud, and lower core-support plate [2,3]. For structural purposes, composites consisting of strong carbon fibers embedded in a carbon matrix are needed. Such carbon/carbon (C/C) composites have been used in aerospace industry to produce missile nose cones, space shuttle leading edge, and aircraft brake-pads. However, radiation-tolerance of such materials is not adequately known because only limited radiation studies have been performed on C/C composites, which suggest that pitch-based carbon fibers have better dimensional stability than that of polyacrylonitrile (PAN) based fibers [4]. The thermodynamically-stable state of graphitic crystalline packing of carbon atoms derived from mesophase pitch leads to a greater stability during neutron irradiation [5]. The specific objectives of this project were: (i) to generating novel carbonaceous nanostructures, (ii) measure extent of graphitic crystallinity and the extent of anisotropy, and (iii) collaborate with the Carbon Materials group at Oak Ridge National Lab to have neutron irradiation studies and post-irradiation examinations conducted on the carbon fibers produced in this research project.« less

Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

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

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

A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooledmore » fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the current state of knowledge is extensive, and in most areas may be sufficient. Several knowledge gaps were identified, such as uncertainty in release from molten fuel and availability of thermodynamic data for lanthanides and actinides in liquid sodium. However, the overall findings suggest that high retention rates can be expected within the fuel and primary sodium for all radionuclides other than noble gases.« less

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

Lubeigt, E.; Laboratoire de Mecanique et d'Acoustique, CNRS UPR 7051, 13402 Marseille Cedex 20; Mensah, S.

The fourth generation of nuclear reactor can use liquid sodium as the core coolant. When the reactor is operating, sodium temperatures can reach up to 600 deg. C. During maintenance periods, when the reactor is shut down, the coolant temperature is reduced to 200 deg. C. Because molten sodium is optically opaque, ultrasonic imaging techniques are developed for maintenance activities. Under-sodium imaging aims at i) checking the health of immersed structures. It should also allow ii) to assess component degradation or damage as cracks and shape defects as well as iii) the detection of lost objects. The under-sodium imaging systemmore » has to sustain high temperature (up to 300 deg. C) and hostility of the sodium environment. Furthermore, specific constraints such as transducers characteristics or the limited sensor mobility in the reactor vessel have to be considered. This work focuses on developing a methodology for detecting damages such as crack defects with ultrasound devices. Surface-breaking cracks or deep cracks are sought in the weld area, as welds are more subject to defects. Traditional methods enabled us to detect emerging cracks of submillimeter size with sodium-compatible high-temperature transducer. The presented approach relies on making use of prior knowledge about the environment through the implementation of differential imaging and time-reversal techniques. Indeed, this approach allows to detect a change by comparison with a reference measurement and by focusing back to any change in the environment. It is a means of analysis and understanding of the physical phenomena making it possible to design more effective inspection strategies. Difference between the measured signals reveals the acoustic field scattered by a perturbation (a crack for instance), which may occur between periodical measurements. The imaging method relies on the adequate combination of two computed ultrasonic fields, one forward and one adjoint. The adjoint field, which carries the information about the defects, is analogous to a time-reversal operation. One of the interests of the presented method is that the time-reversal operation is not done experimentally but numerically. Numerical simulations have been carried out to validate the practical relevance of this approach. The preliminary numerical results show a nice agreement between the guessed and the actual positions of the defect. After water-tests, in sodium-tests must be done in order to validate the water/sodium transposition. For this purpose, an under-sodium device is under development, which can move the transducers with four degrees of freedom in a 1.5 m{sup 3} sodium pot. (authors)« less

The advantages and disadvantages of using the TREAT reactor for nuclear laser experiments

NASA Astrophysics Data System (ADS)

Dickson, P. W.; Snyder, A. M.; Imel, G. R.; McConnell, R. J.

The Transient Reactor Test Facility (TREAT) is a large air-cooled test facility located at the Idaho National Engineering Laboratory. Two of the major design features of TREAT, its large size and its being an air-cooled reactor, provide clues to both its advantages and disadvantages for supporting nuclear laser experiments. Its large size, which is dictated by the dilute uranium/graphite fuel, permits accommodation of geometrically large experiments. However, TREAT's large size also results in relatively long transients so that the energy deposited in an experiment is large relative to the peak power available from the reactor. TREAT's air-cooling mode of operation allows its configuration to be changed fairly readily. Due to air cooling, the reactor cools down slowly, permitting only one full power transient a day, which can be a disadvantage in some experimental programs. The reactor is capable of both steady-state or transient operation.

Teleoperated systems for nuclear reactors: Inspection and maintenance

NASA Technical Reports Server (NTRS)

Dorokhov, V. P.; Dorokhov, D. V.; Eperin, A. P.

1994-01-01

The present paper describes author's work in the field of teleoperated equipment for inspection and maintenance of the RBML technological channels and graphite laying, emergency operations. New technological and design solutions of teleoperated robotic systems developed for Leningradsky Power Plant are discussed.

THE EFFECT OF VOLTAGE ON ELECTROCHEMICAL DEGRADATION OF TRICHLOROETHYLENE

EPA Science Inventory

This study investigates electrochemical degradation of Trichloroethylene (TCE) using granular graphite as electrodes in a flow-through reactor system. The experiments were conducted to obtain information on the effect of voltage and flow rates on the degradation rates of TCE. The...

A school investigation into Chernobyl fallout

NASA Astrophysics Data System (ADS)

Plant, R. D.

1988-01-01

The nuclear power station operating at Chernobyl, just north of Kiev in the Ukraine, USSR, contains four RBMK reactors operating at 1000 MW each. The RBMK reactor is a graphite moderated light water cooled reactor using low enriched uranium fuel. Early on Saturday 26 April 1986 a serious accident occurred to one of the four reactors resulting in the release of radioactive material, some of which was carried by the wind northwards across Poland and Scandinavia. The Ursuline Convent School at Westgate-on-Sea is situated in a small seaside town on the North Kent coast. On 30 April the background count was measured in the physics laboratory of the school using a Mullard ZP1481 Geiger-Muller tube in conjunction with a Panax scaler.

Preparation of Graphene Sheets by Electrochemical Exfoliation of Graphite in Confined Space and Their Application in Transparent Conductive Films.

PubMed

Wang, Hui; Wei, Can; Zhu, Kaiyi; Zhang, Yu; Gong, Chunhong; Guo, Jianhui; Zhang, Jiwei; Yu, Laigui; Zhang, Jingwei

2017-10-04

A novel electrochemical exfoliation mode was established to prepare graphene sheets efficiently with potential applications in transparent conductive films. The graphite electrode was coated with paraffin to keep the electrochemical exfoliation in confined space in the presence of concentrated sodium hydroxide as the electrolyte, yielding ∼100% low-defect (the D band to G band intensity ratio, I D /I G = 0.26) graphene sheets. Furthermore, ozone was first detected with ozone test strips, and the effect of ozone on the exfoliation of graphite foil and the microstructure of the as-prepared graphene sheets was investigated. Findings indicate that upon applying a low voltage (3 V) on the graphite foil partially coated with paraffin wax that the coating can prevent the insufficiently intercalated graphite sheets from prematurely peeling off from the graphite electrode thereby affording few-layer (<5 layers) holey graphene sheets in a yield of as much as 60%. Besides, the ozone generated during the electrochemical exfoliation process plays a crucial role in the exfoliation of graphite, and the amount of defect in the as-prepared graphene sheets is dependent on electrolytic potential and electrode distance. Moreover, the graphene-based transparent conductive films prepared by simple modified vacuum filtration exhibit an excellent transparency and a low sheet resistance after being treated with NH 4 NO 3 and annealing (∼1.21 kΩ/□ at ∼72.4% transmittance).

Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

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

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

1984-06-01

ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Componentmore » Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.« less

Development of a Scanning Microscale Fast Neutron Irradiation Platform for Examining the Correlation Between Local Neutron Damage and Graphite Microstructure

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

Pinhero, Patrick; Windes, William

2015-03-10

The fast particle radiation damage effect of graphite, a main material in current and future nuclear reactors, has significant influence on the utilization of this material in fission and fusion plants. Atoms on graphite crystals can be easily replaced or dislocated by fast protons and result in interstitials and vacancies. The currently accepted model indicates that after most of the interstitials recombine with vacancies, surviving interstitials form clusters and furthermore gather to create loops with each other between layers. Meanwhile, surviving vacancies and interstitials form dislocation loops on the layers. The growth of these inserted layers cause the dimensional increase,more » i.e. swelling, of graphite. Interstitial and vacancy dislocation loops have been reported and they can easily been observed by electron microscope. However, observation of the intermediate atom clusters becomes is paramount in helping prove this model. We utilize fast protons generated from the University of Missouri Research Reactor (MURR) cyclotron to irradiate highly- oriented pyrolytic graphite (HOPG) as target for this research. Post-irradiation examination (PIE) of dosed targets with high-resolution transmission electron microscopy (HRTEM) has permit observation and analysis of clusters and dislocation loops to support the proposed theory. Another part of the research is to validate M.I. Heggie’s Ruck and Tuck model, which introduced graphite layers may fold under fast particle irradiation. Again, we employed microscopy to image irradiated specimens to determine how the extent of Ruck and Tuck by calculating the number of folds as a function of dose. Our most significant accomplishment is the invention of a novel class of high-intensity pure beta-emitters for long-term lightweight batteries. We have filed four invention disclosure records based on the research conducted in this project. These batteries are lightweight because they consist of carbon and tritium and can be fabricated to conform to many geometric shapes. In addition, we have published eight peer-reviewed American Nuclear Society (ANS) transactions, and presented our findings at ANS National Meetings, and several universities.« less

Initial Assessment of X-Ray Computer Tomography image analysis for material defect microstructure

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

Kane, Joshua James; Windes, William Enoch

2016-06-01

The original development work leading to this report was focused on the non destructive three-dimensional (3-D) characterization of nuclear graphite as a means to better understand the nature of the inherent pore structure. The pore structure of graphite and its evolution under various environmental factors such as irradiation, mechanical stress, and oxidation plays an important role in their observed properties and characteristics. If we are to transition from an empirical understanding of graphite behavior to a truly predictive mechanistic understanding the pore structure must be well characterized and understood. As the pore structure within nuclear graphite is highly interconnected andmore » truly 3-D in nature, 3-D characterization techniques are critical. While 3-D characterization has been an excellent tool for graphite pore characterization, it is applicable to a broad number of materials systems over many length scales. Given the wide range of applications and the highly quantitative nature of the tool, it is quite surprising to discover how few materials researchers understand and how valuable of a tool 3-D image processing and analysis can be. Ultimately, this report is intended to encourage broader use of 3 D image processing and analysis in materials science and engineering applications, more specifically nuclear-related materials applications, by providing interested readers with enough familiarity to explore its vast potential in identifying microstructure changes. To encourage this broader use, the report is divided into two main sections. Section 2 provides an overview of some of the key principals and concepts needed to extract a wide variety of quantitative metrics from a 3-D representation of a material microstructure. The discussion includes a brief overview of segmentation methods, connective components, morphological operations, distance transforms, and skeletonization. Section 3 focuses on the application of concepts from Section 2 to relevant materials at Idaho National Laboratory. In this section, image analysis examples featuring nuclear graphite will be discussed in detail. Additionally, example analyses from Transient Reactor Test Facility low-enriched uranium conversion, Advanced Gas Reactor like compacts, and tristructural isotopic particles are shown to give a broader perspective of the applicability to relevant materials of interest.« less

Microwave Diffraction Techniques from Macroscopic Crystal Models

ERIC Educational Resources Information Center

Murray, William Henry

1974-01-01

Discusses the construction of a diffractometer table and four microwave models which are built of styrofoam balls with implanted metallic reflecting spheres and designed to simulate the structures of carbon (graphite structure), sodium chloride, tin oxide, and palladium oxide. Included are samples of Bragg patterns and computer-analysis results.…

Treatment of irradiated graphite from French Bugey reactor

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

Stevens, Howard; Laurent, Gerard

In 2008, following the general French plan for nuclear waste management, Electricite de France attempted to find for irradiated graphite an alternative solution to direct storage at the low-activity long-life storage center in France managed by the national agency for wastes (ANDRA). EDF management requested that its engineering arm, EDF CIDEN, study the graphite treatment alternatives to direct storage. In mid-2008, this study revealed the potential advantage for EDF to use a steam reforming process known as Thermal Organic Reduction, 'THOR' (owned by Studsvik, Inc., USA), to treat or destroy the graphite matrix and limit the quantity of secondary wastemore » to be stored. In late 2009, EDF began a test program with Studsvik to determine if the THOR steam reforming process could be used to destroy the graphite. The program also sought to determine if the graphite could be treated to release the bulk of activity while minimizing the gasification of the bulk mass of the graphite. In October 2009, tests with non-irradiated graphite were completed and demonstrated destruction of a graphite matrix by the THOR process at satisfactory rates. After gasifying the graphite, focus shifted to the effect of roasting graphite at high temperatures in inert gases with low concentrations of oxidizing gases to preferentially remove volatile radionuclides while minimizing the graphite mass loss to 5%. A radioactive graphite sleeve was imported from France to the US for these tests. Completed in April 2010, 'Phase I' of testing showed that the process removed >99% of H-3 and 46% of C-14 with <6% mass loss. Completed in September 2011, 'Phase II' testing achieved increased removals as high as 80% C-14. During Phase II, it was also discovered that roasting in a reducing atmosphere helped to limit the oxidation of the graphite. Future work seeks to explore the effects of reducing gases to limit the bulk oxidation of graphite. If the graphite could be decontaminated of long-lived radionuclides up to 95% for C-14 while minimizing mass loss to <5%, this would minimize the volume of any secondary waste streams and potentially lower the waste class of the larger bulk of graphite. Alternatively, if up to 95% decontamination of C-14 is achieved, the graphite may be completely gasified which could result in lower disposal. (authors)« less

Multi-Physics Simulation of TREAT Kinetics using MAMMOTH

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

DeHart, Mark; Gleicher, Frederick; Ortensi, Javier

With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific fuels transient tests range from simple temperature transients to full fuel melt accidents. The current TREAT core is driven by highly enriched uranium (HEU) dispersed in amore » graphite matrix (1:10000 U-235/C atom ratio). At the center of the core, fuel is removed allowing for the insertion of an experimental test vehicle. TREAT’s design provides experimental flexibility and inherent safety during neutron pulsing. This safety stems from the graphite in the driver fuel having a strong negative temperature coefficient of reactivity resulting from a thermal Maxwellian shift with increased leakage, as well as graphite acting as a temperature sink. Air cooling is available, but is generally used post-transient for heat removal. DOE and INL have expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility, with an emphasis on effective and safe operation while minimizing experimental time and cost. At INL, the Multi-physics Object Oriented Simulation Environment (MOOSE) has been selected as the model development framework for this work. This paper describes the results of preliminary simulations of a TREAT fuel element under transient conditions using the MOOSE-based MAMMOTH reactor physics tool.« less

An Overview of Reactor Concepts, a Survey of Reactor Designs.

DTIC Science & Technology

1985-02-01

may be very different. HTGRs may use highly enriched uranium, thereby yielding better fuel economy and a reduc- tion of the actual core size for a...specific power level. The HTGR core may have fuel and control rods placed in graphite arrays similar to PWR core con- figuration, or they may have fuel ...rods are pulled out. A Peach Bottom core design is another HTGR design. This design is featured by the fuel pin’s ability to purge itself of fission

Microwave plasma CVD of NANO structured tin/carbon composites

DOEpatents

Marcinek, Marek [Warszawa, PL; Kostecki, Robert [Lafayette, CA

2012-07-17

A method for forming a graphitic tin-carbon composite at low temperatures is described. The method involves using microwave radiation to produce a neutral gas plasma in a reactor cell. At least one organo tin precursor material in the reactor cell forms a tin-carbon film on a supporting substrate disposed in the cell under influence of the plasma. The three dimensional carbon matrix material with embedded tin nanoparticles can be used as an electrode in lithium-ion batteries.

Nuclear Data Uncertainty Propagation to Reactivity Coefficients of a Sodium Fast Reactor

NASA Astrophysics Data System (ADS)

Herrero, J. J.; Ochoa, R.; Martínez, J. S.; Díez, C. J.; García-Herranz, N.; Cabellos, O.

2014-04-01

The assessment of the uncertainty levels on the design and safety parameters for the innovative European Sodium Fast Reactor (ESFR) is mandatory. Some of these relevant safety quantities are the Doppler and void reactivity coefficients, whose uncertainties are quantified. Besides, the nuclear reaction data where an improvement will certainly benefit the design accuracy are identified. This work has been performed with the SCALE 6.1 codes suite and its multigroups cross sections library based on ENDF/B-VII.0 evaluation.

Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

NASA Astrophysics Data System (ADS)

Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

2017-09-01

VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

Liquid sodium dip seal maintenance system

DOEpatents

Briggs, Richard L.; Meacham, Sterling A.

1980-01-01

A system for spraying liquid sodium onto impurities associated with liquid dip seals of nuclear reactors. The liquid sodium mixing with the impurities dissolves the impurities in the liquid sodium. The liquid sodium having dissolved and diluted the impurities carries the impurities away from the site thereby cleaning the liquid dip seal and surrounding area. The system also allows wetting of the metallic surfaces of the dip seal thereby reducing migration of radioactive particles across the wetted boundary.

Creation of the NaSCoRD Database

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

Denman, Matthew R.; Jankovsky, Zachary Kyle; Stuart, William

This report was written as part of a United States Department of Energy (DOE), Office of Nuclear Energy, Advanced Reactor Technologies program funded project to re-create the capabilities of the legacy Centralized Reliability Database Organization (CREDO) database. The CREDO database provided a record of component design and performance documentation across various systems that used sodium as a working fluid. Regaining this capability will allow the DOE complex and the domestic sodium reactor industry to better understand how previous systems were designed and built for use in improving the design and operations of future loops. The contents of this report include:more » overview of the current state of domestic sodium reliability databases; summary of the ongoing effort to improve, understand, and process the CREDO information; summary of the initial efforts to develop a unified sodium reliability database called the Sodium System Component Reliability Database (NaSCoRD); and explain both how potential users can access the domestic sodium reliability databases and the type of information that can be accessed from these databases.« less

Verification of Three Dimensional Triangular Prismatic Discrete Ordinates Transport Code ENSEMBLE-TRIZ by Comparison with Monte Carlo Code GMVP

NASA Astrophysics Data System (ADS)

Homma, Yuto; Moriwaki, Hiroyuki; Ohki, Shigeo; Ikeda, Kazumi

2014-06-01

This paper deals with verification of three dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at beginning of cycle of an initial core and at beginning and end of cycle of equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multi-plication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity.

Atmospheric Dispersion of Sodium Aerosol due to a Sodium Leak in a Fast Breeder Reactor Complex

NASA Astrophysics Data System (ADS)

Punitha, G.; Sudha, A. Jasmin; Kasinathan, N.; Rajan, M.

Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model does not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Technical Reports Server (NTRS)

Martin, James; Salvail, Pat

2003-01-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final "wet in". A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/- 1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to less than10(exp -10) std cc/sec helium and vacuum conditioned at 250 C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Sodium Heat Pipe Module Processing For the SAFE-100 Reactor Concept

NASA Astrophysics Data System (ADS)

Martin, James; Salvail, Pat

2004-02-01

To support development and hardware-based testing of various space reactor concepts, the Early Flight Fission-Test Facility (EFF-TF) team established a specialized glove box unit with ancillary systems to handle/process alkali metals. Recently, these systems have been commissioned with sodium supporting the fill of stainless steel heat pipe modules for use with a 100 kW thermal heat pipe reactor design. As part of this effort, procedures were developed and refined to govern each segment of the process covering: fill, leak check, vacuum processing, weld closeout, and final ``wet in''. A series of 316 stainless steel modules, used as precursors to the actual 321 stainless steel modules, were filled with 35 +/-1 grams of sodium using a known volume canister to control the dispensed mass. Each module was leak checked to <10-10 std cc/sec helium and vacuum conditioned at 250 °C to assist in the removal of trapped gases. A welding procedure was developed to close out the fill stem preventing external gases from entering the evacuated module. Finally the completed modules were vacuum fired at 750 °C allowing the sodium to fully wet the internal surface and wick structure of the heat pipe module.

Gap Size Uncertainty Quantification in Advanced Gas Reactor TRISO Fuel Irradiation Experiments

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

Pham, Binh T.; Einerson, Jeffrey J.; Hawkes, Grant L.

The Advanced Gas Reactor (AGR)-3/4 experiment is the combination of the third and fourth tests conducted within the tristructural isotropic fuel development and qualification research program. The AGR-3/4 test consists of twelve independent capsules containing a fuel stack in the center surrounded by three graphite cylinders and shrouded by a stainless steel shell. This capsule design enables temperature control of both the fuel and the graphite rings by varying the neon/helium gas mixture flowing through the four resulting gaps. Knowledge of fuel and graphite temperatures is crucial for establishing the functional relationship between fission product release and irradiation thermal conditions.more » These temperatures are predicted for each capsule using the commercial finite-element heat transfer code ABAQUS. Uncertainty quantification reveals that the gap size uncertainties are among the dominant factors contributing to predicted temperature uncertainty due to high input sensitivity and uncertainty. Gap size uncertainty originates from the fact that all gap sizes vary with time due to dimensional changes of the fuel compacts and three graphite rings caused by extended exposure to high temperatures and fast neutron irradiation. Gap sizes are estimated using as-fabricated dimensional measurements at the start of irradiation and post irradiation examination dimensional measurements at the end of irradiation. Uncertainties in these measurements provide a basis for quantifying gap size uncertainty. However, lack of gap size measurements during irradiation and lack of knowledge about the dimension change rates lead to gap size modeling assumptions, which could increase gap size uncertainty. In addition, the dimensional measurements are performed at room temperature, and must be corrected to account for thermal expansion of the materials at high irradiation temperatures. Uncertainty in the thermal expansion coefficients for the graphite materials used in the AGR-3/4 capsules also increases gap size uncertainty. This study focuses on analysis of modeling assumptions and uncertainty sources to evaluate their impacts on the gap size uncertainty.« less

Innovative technologies on fuel assemblies cleaning for sodium fast reactors: First considerations on cleaning process

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

Simon, N.; Lorcet, H.; Beauchamp, F.

2012-07-01

Within the framework of Sodium Fast Reactor development, innovative fuel assembly cleaning operations are investigated to meet the GEN IV goals of safety and of process development. One of the challenges is to mitigate the Sodium Water Reaction currently used in these processes. The potential applications of aqueous solutions of mineral salts (including the possibility of using redox chemical reactions) to mitigate the Sodium Water Reaction are considered in a first part and a new experimental bench, dedicated to this study, is described. Anhydrous alternative options based on Na/CO{sub 2} interaction are also presented. Then, in a second part, amore » functional study conducted on the cleaning pit is proposed. Based on experimental feedback, some calculations are carried out to estimate the sodium inventory on the fuel elements, and physical methods like hot inert gas sweeping to reduce this inventory are also presented. Finally, the implementation of these innovative solutions in cleaning pits is studied in regard to the expected performances. (authors)« less

Natural precursor based hydrothermal synthesis of sodium carbide for reactor applications

NASA Astrophysics Data System (ADS)

Swapna, M. S.; Saritha Devi, H. V.; Sebastian, Riya; Ambadas, G.; Sankararaman, S.

2017-12-01

Carbides are a class of materials with high mechanical strength and refractory nature which finds a wide range of applications in industries and nuclear reactors. The existing synthesis methods of all types of carbides have problems in terms of use of toxic chemical precursors, high-cost, etc. Sodium carbide (Na2C2) which is an alkali metal carbide is the least explored one and also that there is no report of low-cost and low-temperature synthesis of sodium carbide using the eco-friendly, easily available natural precursors. In the present work, we report a simple low-cost, non-toxic hydrothermal synthesis of refractory sodium carbide using the natural precursor—Pandanus. The formation of sodium carbide along with boron carbide is evidenced by the structural and morphological characterizations. The sample thus synthesized is subjected to field emission scanning electron microscopy (FESEM), x-ray powder diffraction (XRD), ultraviolet (UV)—visible spectroscopy, Fourier transform infrared spectroscopy (FTIR), Raman, and photoluminescent (PL) spectroscopic techniques.

Measurement of cesium diffusion coefficients in graphite IG-110

NASA Astrophysics Data System (ADS)

Carter, L. M.; Brockman, J. D.; Loyalka, S. K.; Robertson, J. D.

2015-05-01

An understanding of the transport of fission products in High Temperature Gas-Cooled Reactors (HTGRs) is needed for operational safety as well as source term estimations. We have measured diffusion coefficients of Cs in IG-110 by using the release method, wherein we infused small graphite spheres with Cs and measured the release rates using ICP-MS. Diffusion behavior was investigated in the temperature range of 1100-1300 K. We have obtained: DCs = (1.0 ×10-7m2 /s) exp(-1.1/×105J /mol RT) and, compared our results with those available in the literature.

A unifying picture of gas-phase formation and growth of PAH (Polycyclic Aromatic Hydrocarbons), soot, diamond and graphite

NASA Technical Reports Server (NTRS)

Frenklach, Michael

1990-01-01

A variety of seemingly different carbon formation processes -- polycyclic aromatic hydrocarbons and diamond in the interstellar medium, soot in hydrocarbon flames, graphite and diamond in plasma-assisted-chemical vapor deposition reactors -- may all have closely related underlying chemical reaction mechanisms. Two distinct mechanisms for gas-phase carbon growth are discussed. At high temperatures it proceeds via the formation of carbon clusters. At lower temperatures it follows a polymerization-type kinetic sequence of chemical reactions of acetylene addition to a radical, and reactivation of the resultant species through H-abstraction by a hydrogen atom.

Apparatus for controlling molten core debris

DOEpatents

Golden, Martin P. [Trafford, PA; Tilbrook, Roger W. [Monroeville, PA; Heylmun, Neal F. [Pittsburgh, PA

1977-07-19

Apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed.

Heterogeneous Reaction gaseous chlorine nitrate and solid sodium chloride

NASA Technical Reports Server (NTRS)

Timonen, Raimo S.; Chu, Liang T.; Leu, Ming-Taun

1994-01-01

The heterogeneous reaction of gaseous chlorine nitrate and solid sodium chloride was investigated over a temperature range of 220 - 300 K in a flow-tube reactor interfaced with a differentially pumped quadrupole mass spectrometer.

Key Assets for a Sustainable Low Carbon Energy Future

NASA Astrophysics Data System (ADS)

Carre, Frank

2011-10-01

Since the beginning of the 21st century, concerns of energy security and climate change gave rise to energy policies focused on energy conservation and diversified low-carbon energy sources. Provided lessons of Fukushima accident are evidently accounted for, nuclear energy will probably be confirmed in most of today's nuclear countries as a low carbon energy source needed to limit imports of oil and gas and to meet fast growing energy needs. Future challenges of nuclear energy are then in three directions: i) enhancing safety performance so as to preclude any long term impact of severe accident outside the site of the plant, even in case of hypothetical external events, ii) full use of Uranium and minimization long lived radioactive waste burden for sustainability, and iii) extension to non-electricity energy products for maximizing the share of low carbon energy source in transportation fuels, industrial process heat and district heating. Advanced LWRs (Gen-III) are today's best available technologies and can somewhat advance nuclear energy in these three directions. However, breakthroughs in sustainability call for fast neutron reactors and closed fuel cycles, and non-electric applications prompt a revival of interest in high temperature reactors for exceeding cogeneration performances achievable with LWRs. Both types of Gen-IV nuclear systems by nature call for technology breakthroughs to surpass LWRs capabilities. Current resumption in France of research on sodium cooled fast neutron reactors (SFRs) definitely aims at significant progress in safety and economic competitiveness compared to earlier reactors of this type in order to progress towards a new generation of commercially viable sodium cooled fast reactor. Along with advancing a new generation of sodium cooled fast reactor, research and development on alternative fast reactor types such as gas or lead-alloy cooled systems (GFR & LFR) is strategic to overcome technical difficulties and/or political opposition specific to sodium. In conclusion, research and technology breakthroughs in nuclear power are needed for shaping a sustainable low carbon future. International cooperation is key for sharing costs of research and development of the required novel technologies and cost of first experimental reactors needed to demonstrate enabling technologies. At the same time technology breakthroughs are developed, pre-normative research is required to support codification work and harmonized regulations that will ultimately apply to safety and security features of resulting innovative reactor types and fuel cycles.

Regulatory Technology Development Plan - Sodium Fast Reactor. Mechanistic Source Term - Metal Fuel Radionuclide Release

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

Grabaskas, David; Bucknor, Matthew; Jerden, James

2016-02-01

The development of an accurate and defensible mechanistic source term will be vital for the future licensing efforts of metal fuel, pool-type sodium fast reactors. To assist in the creation of a comprehensive mechanistic source term, the current effort sought to estimate the release fraction of radionuclides from metal fuel pins to the primary sodium coolant during fuel pin failures at a variety of temperature conditions. These release estimates were based on the findings of an extensive literature search, which reviewed past experimentation and reactor fuel damage accidents. Data sources for each radionuclide of interest were reviewed to establish releasemore » fractions, along with possible release dependencies, and the corresponding uncertainty levels. Although the current knowledge base is substantial, and radionuclide release fractions were established for the elements deemed important for the determination of offsite consequences following a reactor accident, gaps were found pertaining to several radionuclides. First, there is uncertainty regarding the transport behavior of several radionuclides (iodine, barium, strontium, tellurium, and europium) during metal fuel irradiation to high burnup levels. The migration of these radionuclides within the fuel matrix and bond sodium region can greatly affect their release during pin failure incidents. Post-irradiation examination of existing high burnup metal fuel can likely resolve this knowledge gap. Second, data regarding the radionuclide release from molten high burnup metal fuel in sodium is sparse, which makes the assessment of radionuclide release from fuel melting accidents at high fuel burnup levels difficult. This gap could be addressed through fuel melting experimentation with samples from the existing high burnup metal fuel inventory.« less

EFFECTS OF REACTION PARAMETERS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE RATE AND BY-PRODUCTS

EPA Science Inventory

Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas prod...

Development of Electroactive and Anaerobic Ammonium-Oxidizing (Anammox) Biofilms from Digestate in Microbial Fuel Cells.

PubMed

Di Domenico, Enea Gino; Petroni, Gianluca; Mancini, Daniele; Geri, Alberto; Di Palma, Luca; Ascenzioni, Fiorentina

2015-01-01

Microbial Fuel cells (MFCs) have been proposed for nutrient removal and energy recovery from different wastes. In this study the anaerobic digestate was used to feed H-type MFC reactors, one with a graphite anode preconditioned with Geobacter sulfurreducens and the other with an unconditioned graphite anode. The data demonstrate that the digestate acts as a carbon source, and even in the absence of anode preconditioning, electroactive bacteria colonise the anodic chamber, producing a maximum power density of 172.2 mW/m(2). The carbon content was also reduced by up to 60%, while anaerobic ammonium oxidation (anammox) bacteria, which were found in the anodic compartment of the reactors, contributed to nitrogen removal from the digestate. Overall, these results demonstrate that MFCs can be used to recover anammox bacteria from natural sources, and it may represent a promising bioremediation unit in anaerobic digestor plants for the simultaneous nitrogen removal and electricity generation using digestate as substrate.

Development of Electroactive and Anaerobic Ammonium-Oxidizing (Anammox) Biofilms from Digestate in Microbial Fuel Cells

PubMed Central

Petroni, Gianluca; Mancini, Daniele; Geri, Alberto; Palma, Luca Di

2015-01-01

Microbial Fuel cells (MFCs) have been proposed for nutrient removal and energy recovery from different wastes. In this study the anaerobic digestate was used to feed H-type MFC reactors, one with a graphite anode preconditioned with Geobacter sulfurreducens and the other with an unconditioned graphite anode. The data demonstrate that the digestate acts as a carbon source, and even in the absence of anode preconditioning, electroactive bacteria colonise the anodic chamber, producing a maximum power density of 172.2 mW/m2. The carbon content was also reduced by up to 60%, while anaerobic ammonium oxidation (anammox) bacteria, which were found in the anodic compartment of the reactors, contributed to nitrogen removal from the digestate. Overall, these results demonstrate that MFCs can be used to recover anammox bacteria from natural sources, and it may represent a promising bioremediation unit in anaerobic digestor plants for the simultaneous nitrogen removal and electricity generation using digestate as substrate. PMID:26273609

Thermal-hydraulic analysis of N Reactor graphite and shield cooling system performance

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

Low, J.O.; Schmitt, B.E.

1988-02-01

A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a worst-case degraded-core accident that requires the GDCS to mitigate the consequences of the accident. An accident scenario previously developed was designed as the hydrogen-mitigation design-basis accident (HMDBA). Previous HMDBA heat transfer analysis,, using the TRUMP-BD code, was used to define the thermal boundary conditions that the GSDS may bemore » exposed to. These TRUMP/HMDBA analysis results were used to define the bounding operating conditions of the GSCS during the course of an HMDBA transient. Nominal and degraded GSCS scenarios were investigated using RELAP5 within or at the bounds of the HMDBA transient. 10 refs., 42 figs., 10 tabs.« less

Determination of gold and cobalt dopants in advanced materials based on tin oxide by slurry sampling high-resolution continuum source graphite furnace atomic absorption spectrometry

NASA Astrophysics Data System (ADS)

Filatova, Daria G.; Eskina, Vasilina V.; Baranovskaya, Vasilisa B.; Vladimirova, Svetlana A.; Gaskov, Alexander M.; Rumyantseva, Marina N.; Karpov, Yuri A.

2018-02-01

A novel approach is developed for the determination of Co and Au dopants in advanced materials based on tin oxide using high-resolution continuum source graphite furnace atomic absorption spectrometry (HR CS GFAAS) with direct slurry sampling. Sodium carboxylmethylcellulose (Na-CMC) is an effective stabilizer for diluted suspensions. Use Na-CMC allows to transfer the analytes into graphite furnace completely and reproducibly. The relative standard deviation obtained by HR CS GFAAS was not higher than 4%. Accuracy was proven by means inductively coupled plasma mass spectrometry (ICP-MS) in solutions after decomposition as a comparative technique. To determine Au and Co in the volume of SnO2, the acid decomposition conditions (HCl, HF) of the samples were suggested by means of an autoclave in a microwave oven.

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

Gregoire, D.C.; Goltz, D.M.; Chakrabarti, C.L.

Graphite furnace atomic absorption spectrometry (GFAAS) is an insensitive technique for determination of uranium. Experiments were conducted using electrothermal vaporization inductively coupled plasma mass spectrometry to investigate the atomization and vaporization of atomic and molecular uranium species in the graphite furnace. ETV-ICP-MS signals for uranium were observed at temperatures well below the appearance temperature of uranium atoms suggesting the vaporization of molecular uranium oxide at temperatures below 2000{degrees}C. Examination of individual uranium ETV-ICP-MS signals reveals the vaporization of uranium carbide at temperatures above 2600{degrees}C. Chemical modifiers such as 0.2% HF and 0.1% CHF{sub 3} in the argon carrier gas, weremore » ineffective in preventing the formation of uranium carbide at 2700{degrees}C. Vaporization of uranium from a tungsten surface using tungsten foil inserted into the graphite tube prevented the formation of uranium carbide and eliminated the ETV-ICP-MS signal suppression caused by a sodium chloride matrix.« less

Pyrolytic Carbon Nanosheets for Ultrafast and Ultrastable Sodium-Ion Storage.

PubMed

Cho, Se Youn; Kang, Minjee; Choi, Jaewon; Lee, Min Eui; Yoon, Hyeon Ji; Kim, Hae Jin; Leal, Cecilia; Lee, Sungho; Jin, Hyoung-Joon; Yun, Young Soo

2018-04-01

Na-ion cointercalation in the graphite host structure in a glyme-based electrolyte represents a new possibility for using carbon-based materials (CMs) as anodes for Na-ion storage. However, local microstructures and nanoscale morphological features in CMs affect their electrochemical performances; they require intensive studies to achieve high levels of Na-ion storage performances. Here, pyrolytic carbon nanosheets (PCNs) composed of multitudinous graphitic nanocrystals are prepared from renewable bioresources by heating. In particular, PCN-2800 prepared by heating at 2800 °C has a distinctive sp 2 carbon bonding nature, crystalline domain size of ≈44.2 Å, and high electrical conductivity of ≈320 S cm -1 , presenting significantly high rate capability at 600 C (60 A g -1 ) and stable cycling behaviors over 40 000 cycles as an anode for Na-ion storage. The results of this study show the unusual graphitization behaviors of a char-type carbon precursor and exceptionally high rate and cycling performances of the resulting graphitic material, PCN-2800, even surpassing those of supercapacitors. © 2018 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Modeling of irradiated graphite (14)C transfer through engineered barriers of a generic geological repository in crystalline rocks.

PubMed

Poskas, Povilas; Grigaliuniene, Dalia; Narkuniene, Asta; Kilda, Raimondas; Justinavicius, Darius

2016-11-01

There are two RBMK-1500 type graphite moderated reactors at the Ignalina nuclear power plant in Lithuania, and they are under decommissioning now. The graphite cannot be disposed of in a near surface repository, because of large amounts of (14)C. Therefore, disposal of the graphite in a geological repository is a reasonable solution. This study presents evaluation of the (14)C transfer by the groundwater pathway into the geosphere from the irradiated graphite in a generic geological repository in crystalline rocks and demonstration of the role of the different components of the engineered barrier system by performing local sensitivity analysis. The speciation of the released (14)C into organic and inorganic compounds as well as the most recent information on (14)C source term was taken into account. Two alternatives were considered in the analysis: disposal of graphite in containers with encapsulant and without it. It was evaluated that the maximal fractional flux of inorganic (14)C into the geosphere can vary from 10(-11)y(-1) (for non-encapsulated graphite) to 10(-12)y(-1) (for encapsulated graphite) while of organic (14)C it was about 10(-3)y(-1) of its inventory. Such difference demonstrates that investigations on the (14)C inventory and chemical form in which it is released are especially important. The parameter with the highest influence on the maximal flux into the geosphere for inorganic (14)C transfer was the sorption coefficient in the backfill and for organic (14)C transfer - the backfill hydraulic conductivity. Copyright © 2016 Elsevier B.V. All rights reserved.

Modeling Fission Product Sorption in Graphite Structures

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

Szlufarska, Izabela; Morgan, Dane; Allen, Todd

2013-04-08

The goal of this project is to determine changes in adsorption and desorption of fission products to/from nuclear-grade graphite in response to a changing chemical environment. First, the project team will employ principle calculations and thermodynamic analysis to predict stability of fission products on graphite in the presence of structural defects commonly observed in very high- temperature reactor (VHTR) graphites. Desorption rates will be determined as a function of partial pressure of oxygen and iodine, relative humidity, and temperature. They will then carry out experimental characterization to determine the statistical distribution of structural features. This structural information will yield distributionsmore » of binding sites to be used as an input for a sorption model. Sorption isotherms calculated under this project will contribute to understanding of the physical bases of the source terms that are used in higher-level codes that model fission product transport and retention in graphite. The project will include the following tasks: Perform structural characterization of the VHTR graphite to determine crystallographic phases, defect structures and their distribution, volume fraction of coke, and amount of sp2 versus sp3 bonding. This information will be used as guidance for ab initio modeling and as input for sorptivity models; Perform ab initio calculations of binding energies to determine stability of fission products on the different sorption sites present in nuclear graphite microstructures. The project will use density functional theory (DFT) methods to calculate binding energies in vacuum and in oxidizing environments. The team will also calculate stability of iodine complexes with fission products on graphite sorption sites; Model graphite sorption isotherms to quantify concentration of fission products in graphite. The binding energies will be combined with a Langmuir isotherm statistical model to predict the sorbed concentration of fission products on each type of graphite site. The model will include multiple simultaneous adsorbing species, which will allow for competitive adsorption effects between different fission product species and O and OH (for modeling accident conditions).« less

Identifying subassemblies by ultrasound to prevent fuel handling error in sodium fast reactors: First test performed in water

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

Paumel, Kevin; Lhuillier, Christian

2015-07-01

Identifying subassemblies by ultrasound is a method that is being considered to prevent handling errors in sodium fast reactors. It is based on the reading of a code (aligned notches) engraved on the subassembly head by an emitting/receiving ultrasonic sensor. This reading is carried out in sodium with high temperature transducers. The resulting one-dimensional C-scan can be likened to a binary code expressing the subassembly type and number. The first test performed in water investigated two parameters: width and depth of the notches. The code remained legible for notches as thin as 1.6 mm wide. The impact of the depthmore » seems minor in the range under investigation. (authors)« less

SAM Theory Manual

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

Hu, Rui

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

A rechargeable iodine-carbon battery that exploits ion intercalation and iodine redox chemistry.

PubMed

Lu, Ke; Hu, Ziyu; Ma, Jizhen; Ma, Houyi; Dai, Liming; Zhang, Jintao

2017-09-13

Graphitic carbons have been used as conductive supports for developing rechargeable batteries. However, the classic ion intercalation in graphitic carbon has yet to be coupled with extrinsic redox reactions to develop rechargeable batteries. Herein, we demonstrate the preparation of a free-standing, flexible nitrogen and phosphorus co-doped hierarchically porous graphitic carbon for iodine loading by pyrolysis of polyaniline coated cellulose wiper. We find that heteroatoms could provide additional defect sites for encapsulating iodine while the porous carbon skeleton facilitates redox reactions of iodine and ion intercalation. The combination of ion intercalation with redox reactions of iodine allows for developing rechargeable iodine-carbon batteries free from the unsafe lithium/sodium metals, and hence eliminates the long-standing safety issue. The unique architecture of the hierarchically porous graphitic carbon with heteroatom doping not only provides suitable spaces for both iodine encapsulation and cation intercalation but also generates efficient electronic and ionic transport pathways, thus leading to enhanced performance.Carbon-based electrodes able to intercalate Li + and Na + ions have been exploited for high performing energy storage devices. Here, the authors combine the ion intercalation properties of porous graphitic carbons with the redox chemistry of iodine to produce iodine-carbon batteries with high reversible capacities.

Alternate Tritium Production Methods Using A Liquid Lithium Target

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

Wilson, J.

For over 60 years, the Savannah River Site’s primary mission has been the production of tritium. From the beginning, the Savannah River National Laboratory (SRNL) has provided the technical foundation to ensure the successful execution of this critical defense mission. SRNL has developed most of the processes used in the tritium mission and provides the research and development necessary to supply this critical component. This project was executed by first developing reactor models that could be used as a neutron source. In parallel to this development calculations were carried out testing the feasibility of accelerator technologies that could also bemore » used for tritium production. Targets were designed with internal moderating material and optimized target was calculated to be capable of 3000 grams using a 1400 MWt sodium fast reactor, 850 grams using a 400 MWt sodium fast reactor, and 100 grams using a 62 MWt reactor, annually.« less

The Treatment of PPCP-Containing Sewage in an Anoxic/Aerobic Reactor Coupled with a Novel Design of Solid Plain Graphite-Plates Microbial Fuel Cell

PubMed Central

Chang, Yi-Tang; Yang, Chu-Wen; Chang, Yu-Jie; Chang, Ting-Chieh; Wei, Da-Jiun

2014-01-01

Synthetic sewage containing high concentrations of pharmaceuticals and personal care products (PPCPs, mg/L level) was treated using an anoxic/aerobic (A/O) reactor coupled with a microbial fuel cell (MFC) at hydraulic retention time (HRT) of 8 h. A novel design of solid plain graphite plates (SPGRPs) was used for the high surface area biodegradation of the PPCP-containing sewage and for the generation of electricity. The average CODCr and total nitrogen removal efficiencies achieved were 97.20% and 83.75%, respectively. High removal efficiencies of pharmaceuticals, including acetaminophen, ibuprofen, and sulfamethoxazole, were also obtained and ranged from 98.21% to 99.89%. A maximum power density of 532.61 mW/cm2 and a maximum coulombic efficiency of 25.20% were measured for the SPGRP MFC at the anode. Distinct differences in the bacterial community were presented at various locations including the mixed liquor suspended solids and biofilms. The bacterial groups involved in PPCP biodegradation were identified as Dechloromonas spp., Sphingomonas sp., and Pseudomonas aeruginosa. This design, which couples an A/O reactor with a novel design of SPGRP MFC, allows the simultaneous removal of PPCPs and successful electricity production. PMID:25197659

Failure Predictions for VHTR Core Components using a Probabilistic Contiuum Damage Mechanics Model

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

Fok, Alex

2013-10-30

The proposed work addresses the key research need for the development of constitutive models and overall failure models for graphite and high temperature structural materials, with the long-term goal being to maximize the design life of the Next Generation Nuclear Plant (NGNP). To this end, the capability of a Continuum Damage Mechanics (CDM) model, which has been used successfully for modeling fracture of virgin graphite, will be extended as a predictive and design tool for the core components of the very high- temperature reactor (VHTR). Specifically, irradiation and environmental effects pertinent to the VHTR will be incorporated into the modelmore » to allow fracture of graphite and ceramic components under in-reactor conditions to be modeled explicitly using the finite element method. The model uses a combined stress-based and fracture mechanics-based failure criterion, so it can simulate both the initiation and propagation of cracks. Modern imaging techniques, such as x-ray computed tomography and digital image correlation, will be used during material testing to help define the baseline material damage parameters. Monte Carlo analysis will be performed to address inherent variations in material properties, the aim being to reduce the arbitrariness and uncertainties associated with the current statistical approach. The results can potentially contribute to the current development of American Society of Mechanical Engineers (ASME) codes for the design and construction of VHTR core components.« less

AGC-4 Experiment Irradiation Monitoring Data Qualification Interim Report

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

Hull, Laurence Charles

2016-08-01

The Graphite Technology Development Program is running a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. The fourth experiment, Advanced Graphite Creep 4 (AGC 4), began with Advanced Test Reactor (ATR) cycle 157D on May 30, 2015, and has been irradiated for two cycles. The capsule was removed from the reactor after ATR cycle 158A, which ended on January 2, 2016, due to interference with another experiment. Irradiation will resume when the interfering experiment is removed from the reactor. This report documents qualification of AGC 4 experiment irradiation monitoring data for use by themore » Advanced Reactor Technologies (ART) Technology Development Office (TDO) Program for research and development activities required to design and license the first HTR nuclear plant. Qualified data meet the requirements for use as described in the experiment planning and quality assurance documents. Failed data do not meet the requirements and provide no useable information. Trend data may not meet all requirements, but still provide some useable information. Use of Trend data requires assessment of how any deficiencies affect a particular use of the data. All thermocouples (TCs) have functioned throughout the AGC-4 experiment. All temperature data are Qualified for use by the ART TDO Program. Argon, helium, and total gas flow data were within expected ranges and are Qualified for use by the ART TDO Program. Discharge gas line moisture values were consistently low during cycle 157D. At the start of cycle 158A, gas moisture briefly spiked to over 600 ppmv and then declined throughout the cycle. Moisture values are within the measurement range of the instrument and are Qualified for use by the ART TDO Program. Graphite creep specimens were subjected to one of three loads, 393, 491, or 589 lbf. For a brief period during cycle 157D between 12:19 on June 2, 2015 and 08:23 on June 11, 2015 the load cells were wired incorrectly resulting in missing stack load data. Missing stack loads were estimated from measured ram pressures using regression equations developed from the existing data from cycle 157D. Estimated stack loads during this period are considered to be an accurate representation of actual load applied to the stacks. These loads deviate slightly from the planned loads. This deviation does not prevent the data from being Qualified for use, but must be taken into account when analyzing the effect of load on creep. Stack displacement increased consistently throughout the first two cycles with total displacement ranging from 0.4 to 0.8 in. During ATR outages, a set of pneumatic rams raised the stacks of graphite creep specimens to ensure the specimens were not stuck within the test train. This stack raising was performed twice. All stacks were raised successfully each time. The load and displacement data are Qualified for use by the ART TDO Program.« less

Manipulating Adsorption-Insertion Mechanisms in Nanostructured Carbon Materials for High-Efficiency Sodium Ion Storage

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

Qiu, Shen; Xiao, Lifen; Sushko, Maria L.

Hard carbon is one of the most promising anode materials for sodium-ion batteries, but the low coulombic efficiency is still a key barrier. In this paper we synthesized a series of nanostructured hard carbon materials with controlled architectures. Using a combination of in-situ XRD mapping, ex-situ NMR, EPR, electrochemical techniques and simulations, an “adsorption-intercalation” (A-I) mechanism is established for Na ion storage. During the initial stages of Na insertion, Na ions adsorb on the defect sites of hard carbon with a wide adsorption energy distribution, producing a sloping voltage profile. In the second stage, Na ions intercalate into graphitic layersmore » with suitable spacing to form NaCx compounds similar to the Li ion intercalation process in graphite, producing a flat low voltage plateau. The cation intercalation with a flat voltage plateau should be enhanced and the sloping region should be avoided. Guided by this knowledge, non-porous hard carbon material has been developed which has achieved high reversible capacity and coulombic efficiency to fulfill practical application.« less

Methods and codes for neutronic calculations of the MARIA research reactor.

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

Andrzejewski, K.; Kulikowska, T.; Bretscher, M. M.

2002-02-18

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6 x 8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminum. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with the RERTR program. At IAE the package of programs was developedmore » to help its operator in optimization of fuel utilization.« less

MEANS FOR COOLING REACTORS

DOEpatents

Wheeler, J.A.

1957-11-01

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

Development of a novel and cost-effective redox sensor for voltammetric determination of pantoprazole sodium during pharmacokinetic studies

PubMed Central

Khashaba, Pakinaz Y.; Ali, Hassan Refat H.

2017-01-01

A pencil graphite electrode modified with poly (bromocresol green (BCG)) was prepared by electro-polymerization process for the determination of pantoprazole sodium. The surface morphology and structure of poly (BCG) film were characterized by scanning electron microscopy and Fourier transform infrared spectroscopy. The determination of pantoprazole sodium in Britton–Robinson buffer (pH 7.0) was carried out by square wave adsorptive stripping voltammetric technique. Under optimum conditions, the linear response of the peak with concentration of the cited drug was in the range of 6.6–360 × 10−8 M with limit of detection of 2.2 × 10−8 M. Moreover, the poly (BCG)-modified electrode has been successfully applied to determine pantoprazole sodium in tablets, vials and during pharmacokinetic studies. PMID:28878983

Improvement and evaluation of thermal, electrical, sealing and mechanical contacts, and their interface materials

NASA Astrophysics Data System (ADS)

Luo, Xiangcheng

Material contacts, including thermal, electrical, seating (fluid sealing and electromagnetic sealing) and mechanical (pressure) contacts, together with their interface materials, were, evaluated, and in some cases, improved beyond the state of the art. The evaluation involved the use of thermal, electrical and mechanical methods. For thermal contacts, this work evaluated and improved the heat transfer efficiency between two contacting components by developing various thermal interface pastes. Sodium silicate based thermal pastes (with boron nitride particles as the thermally conductive filler) as well as polyethylene glycol (PEG) based thermal pastes were developed and evaluated. The optimum volume fractions of BN in sodium silicate based pastes and PEG based pastes were 16% and 18% respectively. The contribution of Li+ ions to the thermal contact conductance in the PEG-based paste was confirmed. For electrical contacts, the relationship between the mechanical reliability and electrical reliability of solder/copper and silver-epoxy/copper joints was addressed. Mechanical pull-out testing was conducted on solder/copper and silver-epoxy/copper joints, while the contact electrical resistivity was measured. Cleansing of the copper surface was more effective for the reliability of silver-epoxy/copper joint than that of solder/copper joint. For sealing contacts, this work evaluated flexible graphite as an electromagnetic shielding gasket material. Flexible graphite was found to be at least comparable to conductive filled silicone (the state of the art) in terms of the shielding effectiveness. The conformability of flexible graphite with its mating metal surface under repeated compression was characterized by monitoring the contact electrical resistance, as the conformability is important to both electromagnetic scaling and fluid waling using flexible graphite. For mechanical contacts, this work focused on the correlation of the interface structure (such as elastic/plastic deformation, oxidation, strain hardening, passive layer damage, fracture, etc.) with the electrical contact resistance, which was measured in real time for contacts under dynamic compression, thus allowing both reversible and irreversible changes to be observed. The materials studied included metals (carbon steel, stainless steel, aluminum and copper), carbon fiber reinforced polymer-matrix composite (nylon-6), ceramic (mortar) and graphite, due to their relevance to fastening, concrete structures, electric brushes and electrical pressure contacts.

Oxygen-consuming chlor alkali cell configured to minimize peroxide formation

DOEpatents

Chlistunoff, Jerzy B [Los Alamos, NM; Lipp, Ludwig [Brookfield, CT; Gottesfeld, Shimshon [Niskayuna, NY

2006-08-01

Oxygen-consuming zero gap chlor-alkali cell was configured to minimize peroxide formation. The cell included an ion-exchange membrane that divided the cell into an anode chamber including an anode and a cathode chamber including an oxygen gas diffusion cathode. The cathode included a single-piece of electrically conducting graphitized carbon cloth. Catalyst and polytetrafluoroethylene were attached to only one side of the cloth. When the cathode was positioned against the cation exchange membrane with the catalyst side away from the membrane, electrolysis of sodium chloride to chlorine and caustic (sodium hydroxide) proceeded with minimal peroxide formation.

Study of guided wave transmission through complex junction in sodium cooled reactor

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

Elie, Q.; Le Bourdais, F.; Jezzine, K.

2015-07-01

Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presentedmore » in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)« less

Carbide Coatings for Nickel Alloys, Graphite and Carbon/Carbon Composites to be used in Fluoride Salt Valves

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

Nagle, Denis; Zhang, Dajie

2015-10-22

The focus of this research was concerned with developing materials technology that supports the evolution of Generation IV Advanced High Temperature Reactor (AHTR) concepts. Specifically, we investigate refractory carbide coatings for 1) nickel alloys, and 2) commercial carbon-carbon composites (CCCs). Numerous compelling reasons have driven us to focus on carbon and carbide materials. First, unlike metals, the strength and modulus of CCCs increase with rising temperature. Secondly, graphite and carbon composites have been proven effective for resisting highly corrosive fluoride melts such as molten cryolite [Na₃AlF₆] at ~1000°C in aluminum reduction cells. Thirdly, graphite and carbide materials exhibit extraordinary radiationmore » damage tolerance and stability up to 2000°C. Finally, carbides are thermodynamically more stable in liquid fluoride salt than the corresponding metals (i.e. Cr and Zr) found in nickel based alloys.« less

Small Reactor Designs Suitable for Direct Nuclear Thermal Propulsion: Interim Report

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

Bruce G. Schnitzler

Advancement of U.S. scientific, security, and economic interests requires high performance propulsion systems to support missions beyond low Earth orbit. A robust space exploration program will include robotic outer planet and crewed missions to a variety of destinations including the moon, near Earth objects, and eventually Mars. Past studies, in particular those in support of both the Strategic Defense Initiative (SDI) and the Space Exploration Initiative (SEI), have shown nuclear thermal propulsion systems provide superior performance for high mass high propulsive delta-V missions. In NASA's recent Mars Design Reference Architecture (DRA) 5.0 study, nuclear thermal propulsion (NTP) was again selectedmore » over chemical propulsion as the preferred in-space transportation system option for the human exploration of Mars because of its high thrust and high specific impulse ({approx}900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. The recently announced national space policy2 supports the development and use of space nuclear power systems where such systems safely enable or significantly enhance space exploration or operational capabilities. An extensive nuclear thermal rocket technology development effort was conducted under the Rover/NERVA, GE-710 and ANL nuclear rocket programs (1955-1973). Both graphite and refractory metal alloy fuel types were pursued. The primary and significantly larger Rover/NERVA program focused on graphite type fuels. Research, development, and testing of high temperature graphite fuels was conducted. Reactors and engines employing these fuels were designed, built, and ground tested. The GE-710 and ANL programs focused on an alternative ceramic-metallic 'cermet' fuel type consisting of UO2 (or UN) fuel embedded in a refractory metal matrix such as tungsten. The General Electric program examined closed loop concepts for space or terrestrial applications as well as open loop systems for direct nuclear thermal propulsion. Although a number of fast spectrum reactor and engine designs suitable for direct nuclear thermal propulsion were proposed and designed, none were built. This report summarizes status results of evaluations of small nuclear reactor designs suitable for direct nuclear thermal propulsion.« less

Sodium storage and injection system

NASA Technical Reports Server (NTRS)

Keeton, A. R. (Inventor)

1979-01-01

A sodium storage and injection system for delivering atomized liquid sodium to a chemical reactor employed in the production of solar grade silicon is disclosed. The system is adapted to accommodate start-up, shut-down, normal and emergency operations, and is characterized by (1) a jacketed injection nozzle adapted to atomize liquefied sodium and (2) a supply circuit connected to the nozzle for delivering the liquefied sodium. The supply circuit is comprised of a plurality of replaceable sodium containment vessels, a pump interposed between the vessels and the nozzle, and a pressurizing circuit including a source of inert gas connected with the vessels for maintaining the sodium under pressure.

Apparatus for controlling molten core debris. [LMFBR

DOEpatents

Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

1977-07-19

Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures.

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

DOE PAGES

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig; ...

2017-07-10

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Experimental studies on metallic fuel relocation in a single-pin core structure of a sodium-cooled fast reactor

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

Kim, Taeil; Harbaruk, Dzmitry; Gerardi, Craig

Experiments dropping molten uranium into test sections of single fuel pin geometry filled with sodium were conducted to investigate relocation behavior of metallic fuel in the core structures of sodium-cooled fast reactors during a hypothetical core disruptive accident. Metallic uranium was used as a fuel material and HT-9M was used as a fuel cladding material in the experiment in order to accurately mock-up the thermo-physical behavior of the relocation. The fuel cladding failed due to eutectic formation between the uranium and HT-9M for all experiments. The extent of the eutectic formation increased with increasing molten uranium temperature. Voids in themore » relocated fuel were observed for all experiments and were likely formed by sodium boiling in contact with the fuel. In one experiment, numerous fragments of the relocated fuel were found. In conclusion, it could be concluded that the injected metallic uranium fuel was fragmented and dispersed in the narrow coolant channel by sodium boiling« less

Facile synthesis of 3D MnNi-layered double hydroxides (LDH)/graphene composites from directly graphites for pseudocapacitor and their electrochemical analysis

NASA Astrophysics Data System (ADS)

Lee, Ilbok; Jeong, Gyoung Hwa; An, Soyeon; Kim, Sang-Wook; Yoon, Songhun

2018-01-01

Herein, MnNi-layered double hydroxides (LDH) were imbibed within the interlayers of graphene nanosheets. The anionic surfactant, sodium dodecyl sulfate played a role of graphite exfoliator adding interaction with metal cations. Using this process, layered MnNi-LDH-graphene nanocomposite was prepared without formation of graphene oxide. When applied into pseudocapacitor electrode, LDH-graphene with optimal ratio between Mn and Ni exhibited very stable cycle with 90% at 1400 cycles and high energy 47.29 Wh kg-1 at the power density of 7473 W kg-1, which was attributed to highly stable layered LDH structure within conductive graphene layers.

Precipitated Silica from Pumice and Carbon Dioxide Gas (Co2) in Bubble Column Reactor

NASA Astrophysics Data System (ADS)

Dewati, R.; Suprihatin, S.; Sumada, K.; Muljani, S.; Familya, M.; Ariani, S.

2018-01-01

Precipitated silica from silica and carbon dioxide gas has been studied successfully. The source of silica was obtained from pumice stone while precipitation process was carried out with carbon dioxide gas (CO2). The sodium silicate solution was obtained by extracting the silica from pumice stone with sodium hydroxide (NaOH) solution and heated to 100 °C for 1 h. The carbon dioxide gas is injected into the aqueous solution of sodium silicate in a bubble column reactor to form precipitated silica. m2/g. The results indicate that the products obtained are precipitate silica have surface area in the range of 100 - 227 m2/g, silica concentration more than 80%, white in appearance, and silica concentration reached 90% at pH 7.

Chemical Characterization and Removal of C-14 from Irradiated Graphite-12010

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

Cleaver, James; McCrory, Shilo; Smith, Tara E.

2012-07-01

Quantities of irradiated graphite waste are expected to drastically increase, which indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research described here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. Characterization of pre- and post-irradiation graphite was conducted to determine bond type, functionalmore » groups, location and concentration of C-14 and its precursors via the use of surface sensitive characterization techniques. Because most surface C-14 originates from neutron activation of nitrogen, an understanding of nitrogen bonding to graphite may lead to a greater understanding of the formation pathway of C-14. However, no single technique provides a complete picture. Therefore, a portfolio of techniques has been developed, with each technique providing another piece to the puzzle that is the chemical nature of the C-14. Scanning Electron Microscopy (SEM), X-Ray Diffraction (XRD), and Raman Spectroscopy were used to evaluate the morphological features of graphite samples. The concentration, chemical composition, and bonding characteristics of C-14 and its precursors were determined through X-ray Photoelectron Spectroscopy (XPS), Time-of-Flight Secondary Ion Mass Spectrometry (SIMS), and Auger and Energy Dispersive X-ray Analysis Spectroscopy (EDX). High-surface-area graphite foam, POCOFoam{sup R}, was exposed to liquid nitrogen and irradiated. Characterization of this material has shown C-14 to C-12 ratios of 0.035. This information was used to optimize the thermal treatment of graphite. Thermal treatment of irradiated graphite as reported by Fachinger et al. (2007) uses naturally adsorbed oxygen complexes to gasify graphite, thus its effectiveness is highly dependent on the availability of adsorbed oxygen compounds. In research presented, the quantity and form of adsorbed oxygen complexes in pre- and post irradiated graphite was studied using SIMS and XPS. SIMS and XPS detected adsorbed oxygen compounds on both irradiated and unirradiated graphite. During thermal treatment graphite samples are heated in the presence of inert argon gas, which carries off gaseous products released during treatment. Experiments were performed at 900 deg. C and 1400 deg. C to evaluate the selective removal of C-14. (authors)« less

Dual-Layer Oxidation-Protective Plasma-Sprayed SiC-ZrB2/Al2O3-Carbon Nanotube Coating on Graphite

NASA Astrophysics Data System (ADS)

Ariharan, S.; Sengupta, Pradyut; Nisar, Ambreen; Agnihotri, Ankur; Balaji, N.; Aruna, S. T.; Balani, Kantesh

2017-02-01

Graphite is used in high-temperature gas-cooled reactors because of its outstanding irradiation performance and corrosion resistance. To restrict its high-temperature (>873 K) oxidation, atmospheric-plasma-sprayed SiC-ZrB2-Al2O3-carbon nanotube (CNT) dual-layer coating was deposited on graphite substrate in this work. The effect of each layer was isolated by processing each component of the coating via spark plasma sintering followed by isothermal kinetic studies. Based on isothermal analysis and the presence of high residual thermal stress in the oxide scale, degradation appeared to be more severe in composites reinforced with CNTs. To avoid the complexity of analysis of composites, the high-temperature activation energy for oxidation was calculated for the single-phase materials only, yielding values of 11.8, 20.5, 43.5, and 4.5 kJ/mol for graphite, SiC, ZrB2, and CNT, respectively, with increased thermal stability for ZrB2 and SiC. These results were then used to evaluate the oxidation rate for the composites analytically. This study has broad implications for wider use of dual-layer (SiC-ZrB2/Al2O3) coatings for protecting graphite crucibles even at temperatures above 1073 K.

Neutron transmission measurements of poly and pyrolytic graphite crystals

NASA Astrophysics Data System (ADS)

Adib, M.; Abbas, Y.; Abdel-Kawy, A.; Ashry, A.; Kilany, M.; Kenawy, M. A.

The total neutron cross-section measurements of polycrystalline graphite have been carried out in a neutron wavelength from 0.04 to 0.78 nm. This work also presents the neutron transmission measurements of pyrolytic graphite (PG) crystal in a neutron wavelength band from 0.03 to 0.50 nm, at different orientations of the PG crystal with regard to the beam direction. The measurements were performed using three time-of-flight (TOF) spectrometers installed in front of three of the ET-RR-1 reactor horizontal channels. The average value of the coherent scattering amplitude for polycrystalline graphite was calculated and found to be bcoh = (6.61 ± 0.07) fm. The behaviour of neutron transmission through the PG crystal, while oriented at different angles with regard to the beam direction, shows dips at neutron wavelengths corresponding to the reflections from (hkl) planes of hexagonal graphite structure. The positions of the observed dips are found to be in good agreement with the calculated ones. It was also found that a 40 mm thick PG crystal is quite enough to reduce the second-order contamination of the neutron beam from 2.81 to 0.04, assuming that the incident neutrons have a Maxwell distribution with neutron gas temperature 330 K.

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

NASA Astrophysics Data System (ADS)

Lubina, A. S.; Subbotin, A. S.; Sedov, A. A.; Frolov, A. A.

2016-12-01

The fast sodium reactor fuel assembly (FA) with U-Pu-Zr metallic fuel is described. In comparison with a "classical" fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

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

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

In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated meltingmore » of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.« less

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF

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

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset willmore » be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.« less

A Simple Tubular Reactor Experiment.

ERIC Educational Resources Information Center

Hudgins, Robert R.; Cayrol, Bertrand

1981-01-01

Using the hydrolysis of crystal violet dye by sodium hydroxide as an example, the theory, apparatus, and procedure for a laboratory demonstration of tubular reactor behavior are described. The reaction presented can occur at room temperature and features a color change to reinforce measured results. (WB)

Active Range Restoration via Caustic Hydrolysis of Explosively Contaminated Metal Parts

DTIC Science & Technology

2011-05-12

Ammonia, Formate – Sodium Cyanide  Low concentration (40 ppm)  Well established mitigation options • Primary Comp B off-gas components: – Ammonia, NOx...neutralized • In situ sodium cyanide treatment (D/H Reactor) – Hydrogen peroxide (H2O2) oxidizes CN- to CNO- • Analysis of liquid sample for sodium ... cyanide • pH neutralization and hydrolysate thickening – Phosphoric acid (H3PO4) forms buffer – Resulting sodium phosphate thickens to paste Lower

HEDL FACILITIES CATALOG 400 AREA

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

MAYANCSIK BA

1987-03-01

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

The WPI reactor-readying for the next generation

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

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« less

Assessment of Sensor Technologies for Advanced Reactors

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

Korsah, Kofi; Kisner, R. A.; Britton Jr., C. L.

This paper provides an assessment of sensor technologies and a determination of measurement needs for advanced reactors (AdvRx). It is a summary of a study performed to provide the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program. The study covered two broad reactor technology categories: High Temperature Reactors and Fast Reactors. The scope of “High temperature reactors” included Gen IV reactors whose coolant exit temperatures exceed ≈650 °C and are moderated (as opposed to fast reactors). To bound the scope formore » fast reactors, this report reviewed relevant operating experience from US-operated Sodium Fast Reactor (SFR) and relevant test experience from the Fast Flux Test Facility (FFTF). For high temperature reactors the study showed that in many cases instrumentation have performed reasonably well in research and demonstration reactors. However, even in cases where the technology is “mature” (such as thermocouples), HTGRs can benefit from improved technologies. Current HTGR instrumentation is generally based on decades-old technology and adapting newer technologies could provide significant advantages. For sodium fast reactors, the study found that several key research needs arise around (1) radiation-tolerant sensor design for in-vessel or in-core applications, where possible non-invasive sensing approaches for key parameters that minimize the need to deploy sensors in-vessel, (2) approaches to exfiltrating data from in-vessel sensors while minimizing penetrations, (3) calibration of sensors in-situ, and (4) optimizing sensor placements to maximize the information content while minimizing the number of sensors needed.« less

Occurence and prediction of sigma phase in fuel cladding alloys for breeder reactors. [LMFBR

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

Anantatmula, R.P.

1982-01-01

In sodium-cooled fast reactor systems, fuel cladding materials will be exposed for several thousand hours to liquid sodium. Satisfactory performance of the materials depends in part on the sodium compatibility and phase stability of the materials. This paper mainly deals with the phase stability aspect, with particular emphasis on sigma phase formation of the cladding materials upon extended exposures to liquid sodium. A new method of predicting sigma phase formation is proposed for austenitic stainless steels and predictions are compared with the experimental results on fuel cladding materials. Excellent agreement is obtained between theory and experiment. The new method ismore » different from the empirical methods suggested for superalloys and does not suffer from the same drawbacks. The present method uses the Fe-Cr-Ni ternary phase diagram for predicting the sigma-forming tendencies and exhibits a wide range of applicability to austenitic stainless steels and heat-resistant Fe-Cr-Ni alloys.« less

Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

NASA Astrophysics Data System (ADS)

Seiler, J. M.; Rameau, B.

Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

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

Smith, Anthony A.

2013-07-01

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] itmore » is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)« less

Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

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

Brown, Nicholas R.; Pointer, William David; Sieger, Matt

2016-04-01

The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

Application of X-ray micro-CT for micro-structural characterization of APCVD deposited SiC coatings on graphite conduit.

PubMed

Agrawal, A K; Sarkar, P S; Singh, B; Kashyap, Y S; Rao, P T; Sinha, A

2016-02-01

SiC coatings are commonly used as oxidation protective materials in high-temperature applications. The operational performance of the coating depends on its microstructure and uniformity. This study explores the feasibility of applying tabletop X-ray micro-CT for the micro-structural characterization of SiC coating. The coating is deposited over the internal surface of pipe structured graphite fuel tube, which is a prototype of potential components of compact high-temperature reactor (CHTR). The coating is deposited using atmospheric pressure chemical vapor deposition (APCVD) and properties such as morphology, porosity, thickness variation are evaluated. Micro-structural differences in the coating caused by substrate distance from precursor inlet in a CVD reactor are also studied. The study finds micro-CT a potential tool for characterization of SiC coating during its future course of engineering. We show that depletion of reactants at larger distances causes development of larger pores in the coating, which affects its morphology, density and thickness. Copyright © 2015 Elsevier Ltd. All rights reserved.

Radiation Characterization Summary: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Located in the Central Cavity on the 32-Inch Pedestal at the Core Centerline (ACRR-PLG-CC-32-cl).

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

Parma, Edward J.,; Vehar, David W.; Lippert, Lance L.

2015-06-01

This document presents the facility-recommended characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the Annular Core Research Reactor (ACRR) for the polyethylene-lead-graphite (PLG) bucket in the central cavity on the 32-inch pedestal at the core centerline. The designation for this environment is ACRR-PLG-CC-32-cl. The neutron, prompt gamma-ray, and delayed gamma-ray energy spectra, uncertainties, and covariance matrices are presented as well as radial and axial neutron and gamma-ray fluence profiles within the experiment area of the bucket. Recommended constants are given to facilitate the conversion of various dosimetry readings into radiation metrics desired by experimenters. Representative pulsemore » operations are presented with conversion examples. Acknowledgements The authors wish to thank the Annular Core Research Reactor staff and the Radiation Metrology Laboratory staff for their support of this work. Also thanks to David Ames for his assistance in running MCNP on the Sandia parallel machines.« less

On the Solidification and Structure Formation during Casting of Large Inserts in Ferritic Nodular Cast Iron

NASA Astrophysics Data System (ADS)

Tadesse, Abel; Fredriksson, Hasse

2018-06-01

The graphite nodule count and size distributions for boiling water reactor (BWR) and pressurized water reactor (PWR) inserts were investigated by taking samples at heights of 2160 and 1150 mm, respectively. In each cross section, two locations were taken into consideration for both the microstructural and solidification modeling. The numerical solidification modeling was performed in a two-dimensional model by considering the nucleation and growth in eutectic ductile cast iron. The microstructural results reveal that the nodule size and count distribution along the cross sections are different in each location for both inserts. Finer graphite nodules appear in the thinner sections and close to the mold walls. The coarser nodules are distributed mostly in the last solidified location. The simulation result indicates that the finer nodules are related to a higher cooling rate and a lower degree of microsegregation, whereas the coarser nodules are related to a lower cooling rate and a higher degree of microsegregation. The solidification time interval and the last solidifying locations in the BWR and PWR are also different.

CHAIN REACTING SYSTEM

DOEpatents

Fermi, E.; Leverett, M.C.

1958-06-01

A nuclear reactor of the gas-cooled, graphitemoderated type is described. In this design, graphite blocks are arranged in a substantially cylindrical lattice having vertically orienied coolant channels in which uranium fuel elements having through passages are disposed. The active lattice is contained within a hollow body. such as a steel shell, which, in turn, is surrounded by water and concrete shields. Helium is used as the primary coolant and is circulated under pressure through the coolant channels and fuel elements. The helium is then conveyed to heat exchangers, where its heat is used to produce steam for driving a prime mover, thence to filtering means where radioactive impurities are removed. From the filtering means the helium passes to a compressor and an after cooler and is ultimately returned to the reactor for recirculation. Control and safety rods are provided to stabilize or stop the reaction. A space is provided between the graphite lattice and the internal walls of the shell to allow for thermal expansion of the lattice during operation. This space is filled with a resilient packing, such as asbestos, to prevent the passage of helium.

ZnO/graphite composites and its antibacterial activity at different conditions.

PubMed

Dědková, Kateřina; Janíková, Barbora; Matějová, Kateřina; Čabanová, Kristina; Váňa, Rostislav; Kalup, Aleš; Hundáková, Marianna; Kukutschová, Jana

2015-10-01

The paper reports laboratory preparation, characterization and in vitro evaluation of antibacterial activity of ZnO/graphite nanocomposites. Zinc chloride and sodium carbonate served as precursors for synthesis of zinc oxide, while micromilled and natural graphite were used as the matrix for ZnO nanoparticles anchoring. During the reaction of ZnCl2 with saturated aqueous solution of Na2CO3a new compound is created. During the calcination at the temperature of 500 °C this new precursors decomposes and ZnO nanoparticles are formed. Composites ZnO/graphite with 50 wt.% of ZnO particles were prepared. X-ray powder diffraction and Raman microspectroscopy served as phase-analytical methods. Scanning electron microscopy technique was used for morphology characterization of the prepared samples and EDS mapping for visualization of elemental distribution. A developed modification of the standard microdilution test was used for in vitro evaluation of daylight induced antibacterial activity and antibacterial activity at dark conditions. Common human pathogens served as microorganism for antibacterial assay. Antibacterial activity of ZnO/graphite composites could be based on photocatalytic reaction; however there is a role of Zn(2+) ions on the resulting antibacterial activity which proved the experiments in dark condition. There is synergistic effect between Zn(2+) caused and reactive oxygen species caused antibacterial activity. Copyright © 2015 Elsevier B.V. All rights reserved.

METHOD FOR REMOVING SODIUM OXIDE FROM LIQUID SODIUM

DOEpatents

Bruggeman, W.H.; Voorhees, B.G.

1957-12-01

A method is described for removing sodium oxide from a fluent stream of liquid sodium by coldtrapping the sodium oxide. Apparatus utilizing this method is disclosed in United States Patent No. 2,745,552. Sodium will remain in a molten state at temperatures below that at which sodium oxide will crystallize out and form solid deposits, therefore, the contaminated stream of sodium is cooled to a temperature at which the solubility of sodium oxide in sodium is substantially decreased. Thereafter the stream of sodium is passed through a bed of stainless steel wool maintained at a temperature below that of the stream. The stream is kept in contact with the wool until the sodium oxide is removed by crystal growth on the wool, then the stream is reheated and returned to the system. This method is useful in purifying reactor coolants where the sodium oxide would otherwise deposit out on the walls and eventually plug the coolant tubes.

Orientation of surfactant self-assembled aggregates on graphite

NASA Astrophysics Data System (ADS)

Sammalkorpi, Maria; Hynninen, Antti-Pekka; Panagiotopoulos, Athanassios Z.; Haataja, Mikko

2007-03-01

Micellar aggregates on surfaces can provide a self-healing corrosion protection or lubrication layer. It has been observed experimentally that on a single crystal surface this layer often consists of oriented hemi-cylindrical micelles which are aligned with the underlying crystal lattice (``orientation effect''). A key feature of this self-assembly process is the interplay between detergent--detergent and detergent--surface interactions. Since the dimensions of the detergent molecules and the unit cell of the surface are typically quite different, the origins of this orientation effect remain unclear. Here we address the question and present the results of Molecular Dynamics simulations of sodium dodecyl sulfate (SDS) self-aggregation on graphite. We employ both single-molecule and multi-molecule simulations of SDS to unravel the origins of the orientation effect. We report that the underlying graphite surface is sufficient to impose orientational bias on individual SDS molecules diffusing on the surface. This produces collective effects that give rise to the oriented hemi-micelles.

A solid reactor core thermal model for nuclear thermal rockets

NASA Astrophysics Data System (ADS)

Rider, William J.; Cappiello, Michael W.; Liles, Dennis R.

1991-01-01

A Helium/Hydrogen Cooled Reactor Analysis (HERA) computer code has been developed. HERA has the ability to model arbitrary geometries in three dimensions, which allows the user to easily analyze reactor cores constructed of prismatic graphite elements. The code accounts for heat generation in the fuel, control rods, and other structures; conduction and radiation across gaps; convection to the coolant; and a variety of boundary conditions. The numerical solution scheme has been optimized for vector computers, making long transient analyses economical. Time integration is either explicit or implicit, which allows the use of the model to accurately calculate both short- or long-term transients with an efficient use of computer time. Both the basic spatial and temporal integration schemes have been benchmarked against analytical solutions.

Continuous Heterogeneous Photocatalysis in Serial Micro-Batch Reactors.

PubMed

Pieber, Bartholomäus; Shalom, Menny; Antonietti, Markus; Seeberger, Peter H; Gilmore, Kerry

2018-01-29

Solid reagents, leaching catalysts, and heterogeneous photocatalysts are commonly employed in batch processes but are ill-suited for continuous-flow chemistry. Heterogeneous catalysts for thermal reactions are typically used in packed-bed reactors, which cannot be penetrated by light and thus are not suitable for photocatalytic reactions involving solids. We demonstrate that serial micro-batch reactors (SMBRs) allow for the continuous utilization of solid materials together with liquids and gases in flow. This technology was utilized to develop selective and efficient fluorination reactions using a modified graphitic carbon nitride heterogeneous catalyst instead of costly homogeneous metal polypyridyl complexes. The merger of this inexpensive, recyclable catalyst and the SMBR approach enables sustainable and scalable photocatalysis. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

DTIC Science & Technology

2006-12-01

24 Table 3.3 Hazards of Sodium Reaction Products, Hydride And Oxide...........................26 Table 3.4 Chemical Reactivity Of Selected...Liquid Metal Fast Breeder Reactor ORIGEN Oak Ridge Isotope Generator ORIGENARP Oak Ridge Isotope Generator Automated Rapid Processing PWR ...nuclear reactors, both because of the possibility of increased reactivity due to boiling and the potential loss of effectiveness of coolant heat transfer

Rapid solar-thermal decarbonization of methane

NASA Astrophysics Data System (ADS)

Dahl, Jaimee Kristen

Due to the ever-increasing demand for energy and the concern over the environmental impact of continuing to produce energy using current methods, there is interest in developing a hydrogen economy. Hydrogen is a desirable energy source because it is abundant in nature and burns cleanly. One method for producing hydrogen is to utilize a renewable energy source to obtain high enough temperatures to decompose a fossil fuel into its elements. This thesis work is directed at developing a solar-thermal aerosol flow reactor to dissociate methane to carbon black and hydrogen. The technology is intended as a "bridge" between current hydrogen production methods, such as conventional steam-methane reformers, and future "zero emission" technology for producing hydrogen, such as dissociating water using a renewable heating source. A solar furnace is used to heat a reactor to temperatures in excess of 2000 K. The final reactor design studied consists of three concentric vertical tubes---an outer quartz protection tube, a middle solid graphite heating tube, and an inner porous graphite reaction tube. A "fluid-wall" is created on the inside wall of the porous reaction tube in order to prevent deposition of the carbon black co-product on the reactor tube wall. The amorphous carbon black produced aids in heating the gas stream by absorbing radiation from the reactor wall. Conversions of 90% are obtained at a reactor wall temperature of 2100 K and an average residence time of 0.01 s. Computer modeling is also performed to study the gas flow and temperature profiles in the reactor as well as the kinetics of the methane dissociation reaction. The simulations indicate that there is little flow of the fluid-wall gas through the porous wall in the hot zone region, but this can be remedied by increasing the inlet temperature of the fluid-wall gas and/or increasing the tube permeability only in the hot zone region of the wall. The following expression describes the kinetics of methane dissociation in a solar-thermal fluid-wall reactor: dXdt=5.8x108 exp-155,600RT 1-X 7.2s-1. The experimental and theoretical work reported in this thesis is the groundwork that will be utilized in scaling up the reactor to produce hydrogen in distributed or centralized facilities.

Temperature Resistant Fiber Bragg Gratings for On-Line and Structural Health Monitoring of the Next-Generation of Nuclear Reactors.

PubMed

Laffont, Guillaume; Cotillard, Romain; Roussel, Nicolas; Desmarchelier, Rudy; Rougeault, Stéphane

2018-06-02

The harsh environment associated with the next generation of nuclear reactors is a great challenge facing all new sensing technologies to be deployed for on-line monitoring purposes and for the implantation of SHM methods. Sensors able to resist sustained periods at very high temperatures continuously as is the case within sodium-cooled fast reactors require specific developments and evaluations. Among the diversity of optical fiber sensing technologies, temperature resistant fiber Bragg gratings are increasingly being considered for the instrumentation of future nuclear power plants, especially for components exposed to high temperature and high radiation levels. Research programs are supporting the developments of optical fiber sensors under mixed high temperature and radiative environments leading to significant increase in term of maturity. This paper details the development of temperature-resistant wavelength-multiplexed fiber Bragg gratings for temperature and strain measurements and their characterization for on-line monitoring into the liquid sodium used as a coolant for the next generation of fast reactors.

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

DOE PAGES

Bess, John D.; Fujimoto, Nozomu

2014-10-09

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

Novel duplex vapor: Electrochemical method for silicon solar cells. [chemical reactor for a silicon sodium reaction system

NASA Technical Reports Server (NTRS)

Nanis, L.; Sanjurjo, A.; Sancier, K.

1979-01-01

The scaled up chemical reactor for a SiF4-Na reaction system is examined for increased reaction rate and production rate. The reaction system which now produces 5 kg batches of mixed Si and NaF is evaluated. The reactor design is described along with an analysis of the increased capacity of the Na chip feeder. The reactor procedure is discussed and Si coalescence in the reaction products is diagnosed.

Temperature Dependence of Power Reflectivity of the First-Wall Materials in the Synchrotron Radiation Range

NASA Astrophysics Data System (ADS)

Takada, Noriharu; Nagatsu, Masaaki; Shimada, Michiya

1995-07-01

The temperature dependence of power reflectivity in the synchrotron radiation range was measured for candidate first-wall materials of the fusion reactor, such as B4C-coated isotropic graphite, C/C composite material, silicon carbide (SiC), tungsten (W), molybdenum (Mo) and SUS-316. The measurements were carried out using a vacuum vessel with a pressure of about 3 mTorr to avoid oxidation. Distinct temperature dependence of reflectivity was observed only for B4C-coated isotropic graphite. For the other materials, power reflectivities were insensitive to temperature in the range from 300 K to ˜900 K. Theoretical analysis of the results is also presented.

Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle

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

Lubina, A. S., E-mail: lubina-as@nrcki.ru; Subbotin, A. S.; Sedov, A. A.

2016-12-15

The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The resultsmore » of the hydrodynamics and heat transfer calculations have been analyzed.« less

Electrochemistry of Cytochrome P450 BM3 in Sodium Dodecyl Sulfate Films

PubMed Central

Udit, Andrew K.; Hill, Michael G.; Gray, Harry B.

2008-01-01

Direct electrochemistry of the cytochrome P450 BM3 heme domain (BM3) was achieved by confining the protein within sodium dodecyl sulfate (SDS) films on the surface of basal-plane graphite (BPG) electrodes. Cyclic voltammetry revealed the heme FeIII/II redox couple at −330 mV (vs. Ag/AgCl, pH 7.4). Up to 10 V/s, the peak current was linear with scan rate, allowing us to treat the system as surface-confined within this regime. The standard heterogeneous rate constant determined at 10 V/s was estimated to be 10 s−1. Voltammograms obtained for the BM3-SDS-BPG system in the presence of dioxygen exhibited catalytic waves at the onset of FeIII reduction. The altered heme reduction potential of the BM3-SDS-graphite system indicates that SDS is likely bound in the enzyme active-site region. Compared to other P450-surfactant systems, we find redox potentials and electron transfer rates that differ by ~ 100 mV and > 10-fold, respectively, indicating that the nature of the surfactant environment has a significant effect on the observed heme redox properties. PMID:17129070

Initial conceptual design study of self-critical nuclear pumped laser systems

NASA Technical Reports Server (NTRS)

Rodgers, R. J.

1979-01-01

An analytical study of self-critical nuclear pumped laser system concepts was performed. Primary emphasis was placed on reactor concepts employing gaseous uranium hexafluoride (UF6) as the fissionable material. Relationships were developed between the key reactor design parameters including reactor power level, critical mass, neutron flux level, reactor size, operating pressure, and UF6 optical properties. The results were used to select a reference conceptual laser system configuration. In the reference configuration, the 3.2 m cubed lasing volume is surrounded by a graphite internal moderator and a region of heavy water. Results of neutronics calculations yield a critical mass of 4.9 U(235) in the form (235)UF6. The configuration appears capable of operating in a continuous steady-state mode. The average gas temperature in the core is 600 K and the UF6 partial pressure within the lasing volume is 0.34 atm.

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

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

Waris, Abdul, E-mail: awaris@fi.itb.ac.id; Novitrian,; Pramuditya, Syeilendra

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Themore » result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.« less

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.

High Temperature Carbonized Grass as a High Performance Sodium Ion Battery Anode.

PubMed

Zhang, Fang; Yao, Yonggang; Wan, Jiayu; Henderson, Doug; Zhang, Xiaogang; Hu, Liangbing

2017-01-11

Hard carbon is currently considered the most promising anode candidate for room temperature sodium ion batteries because of its relatively high capacity, low cost, and good scalability. In this work, switchgrass as a biomass example was carbonized under an ultrahigh temperature, 2050 °C, induced by Joule heating to create hard carbon anodes for sodium ion batteries. Switchgrass derived carbon materials intrinsically inherit its three-dimensional porous hierarchical architecture, with an average interlayer spacing of 0.376 nm. The larger interlayer spacing than that of graphite allows for the significant Na ion storage performance. Compared to the sample carbonized under 1000 °C, switchgrass derived carbon at 2050 °C induced an improved initial Coulombic efficiency. Additionally, excellent rate capability and superior cycling performance are demonstrated for the switchgrass derived carbon due to the unique high temperature treatment.

Risk Management for Sodium Fast Reactors.

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

Denman, Matthew R.; Groth, Katrina; Cardoni, Jeffrey N.

2015-01-01

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

Enrichment of acetogenic bacteria in high rate anaerobic reactors under mesophilic and thermophilic conditions.

PubMed

Ryan, P; Forbes, C; McHugh, S; O'Reilly, C; Fleming, G T A; Colleran, E

2010-07-01

The objective of the current study was to expand the knowledge of the role of acetogenic Bacteria in high rate anaerobic digesters. To this end, acetogens were enriched by supplying a variety of acetogenic growth supportive substrates to two laboratory scale high rate upflow anaerobic sludge bed (UASB) reactors operated at 37 degrees C (R1) and 55 degrees C (R2). The reactors were initially fed a glucose/acetate influent. Having achieved high operational performance and granular sludge development and activity, both reactors were changed to homoacetogenic bacterial substrates on day 373 of the trial. The reactors were initially fed with sodium vanillate as a sole substrate. Although % COD removal indicated that the 55 degrees C reactor out performed the 37 degrees C reactor, effluent acetate levels from R2 were generally higher than from R1, reaching values as high as 5023 mg l(-1). Homoacetogenic activity in both reactors was confirmed on day 419 by specific acetogenic activity (SAA) measurement, with higher values obtained for R2 than R1. Sodium formate was introduced as sole substrate to both reactors on day 464. It was found that formate supported acetogenic activity at both temperatures. By the end of the trial, no specific methanogenic activity (SMA) was observed against acetate and propionate indicating that the methane produced was solely by hydrogenotrophic Archaea. Higher SMA and SAA values against H(2)/CO(2) suggested development of a formate utilising acetogenic population growing in syntrophy with hydrogenotrophic methanogens. Throughout the formate trial, the mesophilic reactor performed better overall than the thermophilic reactor. Copyright 2010 Elsevier Ltd. All rights reserved.

The Impact of Alkaliphilic Biofilm Formation on the Release and Retention of Carbon Isotopes from Nuclear Reactor Graphite.

PubMed

Rout, S P; Payne, L; Walker, S; Scott, T; Heard, P; Eccles, H; Bond, G; Shah, P; Bills, P; Jackson, B R; Boxall, S A; Laws, A P; Charles, C; Williams, S J; Humphreys, P N

2018-03-13

14 C is an important consideration within safety assessments for proposed geological disposal facilities for radioactive wastes, since it is capable of re-entering the biosphere through the generation of 14 C bearing gases. The irradiation of graphite moderators in the UK gas-cooled nuclear power stations has led to the generation of a significant volume of 14 C-containing intermediate level wastes. Some of this 14 C is present as a carbonaceous deposit on channel wall surfaces. Within this study, the potential of biofilm growth upon irradiated and 13 C doped graphite at alkaline pH was investigated. Complex biofilms were established on both active and simulant samples. High throughput sequencing showed the biofilms to be dominated by Alcaligenes sp at pH 9.5 and Dietzia sp at pH 11.0. Surface characterisation revealed that the biofilms were limited to growth upon the graphite surface with no penetration of the deeper porosity. Biofilm formation resulted in the generation of a low porosity surface layer without the removal or modification of the surface deposits or the release of the associated 14 C/ 13 C. Our results indicated that biofilm formation upon irradiated graphite is likely to occur at the pH values studied, without any additional release of the associated 14 C.

ETR BUILDING, TRA642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS ...

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

ETR BUILDING, TRA-642, INTERIOR. BASEMENT. ROLLUP DOOR TO CUBICLE POSTS CAUTION SIGNS BECAUSE OF SODIUM HAZARD WITHIN. INL NEGATIVE NO. HD24-3-1. Mike Crane, Photographer, 11/2000 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

DOEpatents

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

1962-06-26

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

Reactor component automatic grapple

DOEpatents

Greenaway, Paul R.

1982-01-01

A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment.

Treatment of Irradiated Graphite from French Bugey Reactor - 13424

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

Brown, Thomas; Poncet, Bernard

2013-07-01

Beginning in 2009, in order to determine an alternative to direct disposal for decommissioned irradiated graphite from EDF's Bugey NPP, Studsvik and EDF began a test program to determine if graphite decontamination and destruction were practicable using Studsvik's thermal organic reduction (THOR) technology. The testing program focused primarily on the release of C-14, H-3, and Cl-36 and also monitored graphite mass loss. For said testing, a bench-scale steam reformer (BSSR) was constructed with the capability of flowing various compositions of gases at temperatures up to 1300 deg. C over uniformly sized particles of graphite for fixed amounts of time. Themore » BSSR was followed by a condenser, thermal oxidizer, and NaOH bubbler system designed to capture H-3 and C-14. Also, in a separate series of testing, high concentration acid and peroxide solutions were used to soak the graphite and leach out and measure Cl-36. A series of gasification tests were performed to scope gas compositions and temperatures for graphite gasification using steam and oxygen. Results suggested higher temperature steam (1100 deg. C vs. 900 deg. C) yielded a practicable gasification rate but that lower temperature (900 deg. C) gasification was also a practicable treatment alternative if oxygen is fed into the process. A series of decontamination tests were performed to determine the release behavior of and extent to which C-14 and H-3 were released from graphite in a high temperature (900-1300 deg. C), low flow roasting gas environment. In general, testing determined that higher temperatures and longer roasting times were efficacious for releasing H-3 completely and the majority (80%) of C-14. Manipulating oxidizing and reducing gas environments was also found to limit graphite mass loss. A series of soaking tests was performed to measure the amount of Cl-36 in the samples of graphite before and after roasting in the BSSR. Similar to C-14 release, these soaking tests revealed that 70-80% Cl-36 is released during roasting tests. (authors)« less

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

NASA Astrophysics Data System (ADS)

Acır, Adem; Altunok, Taner

2010-10-01

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

Interim Status Report for Risk Management for SFRs

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

Jankovsky, Zachary Kyle; Denman, Matthew R.; Groth, Katrina

2015-10-01

Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of passive, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to take advantage of natural phenomena to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a variety of beyondmore » design basis events with the intent of exploring the utility of a Dynamic Bayesian Network to infer the state of the reactor to inform the operator's corrective actions. These inferences also serve to identify the instruments most critical to informing an operator's actions as candidates for hardening against radiation and other extreme environmental conditions that may exist in an accident. This reduction in uncertainty serves to inform ongoing discussions of how small sodium reactors would be licensed and may serve to reduce regulatory risk and cost for such reactors.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

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

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Generating Aromatics From CO2 on Mars or Natural Gas on Earth

NASA Technical Reports Server (NTRS)

Muscatello, Anthony C.; Zubrin, Robert; Berggren, Mark

2006-01-01

Methane to aromatics on Mars ( METAMARS ) is the name of a process originally intended as a means of converting Martian atmospheric carbon dioxide to aromatic hydrocarbons and oxygen, which would be used as propellants for spacecraft to return to Earth. The process has been demonstrated on Earth on a laboratory scale. A truncated version of the process could be used on Earth to convert natural gas to aromatic hydrocarbon liquids. The greater (relative to natural gas) density of aromatic hydrocarbon liquids makes it more economically feasible to ship them to distant markets. Hence, this process makes it feasible to exploit some reserves of natural gas that, heretofore, have been considered as being "stranded" too far from markets to be of economic value. In the full version of METAMARS, carbon dioxide is frozen out of the atmosphere and fed to a Sabatier reactor along with hydrogen (which, on Mars, would have been brought from Earth). In the Sabatier reactor, these feedstocks are converted to methane and water. The water is condensed and electrolyzed to oxygen (which is liquefied) and hydrogen (which is recycled to the Sabatier reactor). The methane is sent to an aromatization reactor, wherein, over a molybdenum-on-zeolite catalyst at a temperature 700 C, it is partially converted into aromatic hydrocarbons (specifically, benzene, toluene, and naphthalene) along with hydrogen. The aromatics are collected by freezing, while unreacted methane and hydrogen are separated by a membrane. Most of the hydrogen is recycled to the Sabatier reactor, while the methane and a small portion of the hydrogen are recycled to the aromatization reactor. The partial recycle of hydrogen to the aromatization reactor greatly increases the catalyst lifetime and eases its regeneration by preventing the formation of graphitic carbon, which could damage the catalyst. (Moreover, if graphitic carbon were allowed to form, it would be necessary to use oxygen to remove it.) Because the aromatics contain only one hydrogen atom per carbon atom, METAMARS produces four times as much propellant from a given amount of hydrogen as does a related process that includes the Sabatier reaction and electrolysis but not aromatization. In the terrestrial version of METAMARS, the Sabatier reactor and electrolyzer would be omitted, while the hydrogen/ methane membrane-separating membrane, the aromatization reactor, and the unreacted-gas-recycling subsystem would be retained. Natural gas would be fed directly to the aromatization reactor. Because natural gas consists of higher hydrocarbons in addition to methane, the aromatization subprocess should be more efficient than it is for methane alone.

Various methods to improve heat transfer in exchangers

NASA Astrophysics Data System (ADS)

Pavel, Zitek; Vaclav, Valenta

2015-05-01

The University of West Bohemia in Pilsen (Department of Power System Engineering) is working on the selection of effective heat exchangers. Conventional shell and tube heat exchangers use simple segmental baffles. It can be replaced by helical baffles, which increase the heat transfer efficiency and reduce pressure losses. Their usage is demonstrated in the primary circuit of IV. generation MSR (Molten Salt Reactors). For high-temperature reactors we consider the use of compact desk heat exchangers, which are small, which allows the integral configuration of reactor. We design them from graphite composites, which allow up to 1000°C and are usable as exchangers: salt-salt or salt-acid (e.g. for the hydrogen production). In the paper there are shown thermo-physical properties of salts, material properties and principles of calculations.

Sodium-22 In Supernovae

NASA Astrophysics Data System (ADS)

Amari, Sachiko

2008-05-01

There are several isotopically distinct noble gas components in meteorites. Of them, Ne-E(L), heavily enriched in 22Ne, is carried by graphite with a range of density (1.6 - 2.2 g/cm3). Bulk (=aggregates) noble gas analysis of graphite separates from the Murchison meteorite indicate that a dominant source of 22Ne is 22Na (T1/2 = 2.6 a) with varying proportions of 22Ne via 14N(α,γ)18F(e+ν)18O(α,γ)22Ne with density. Low-density graphite grains, from their isotopic signatures, are believed to have formed in supernovae. Examinations of both bulk and single-grain analyses of low-density graphite grains (Amari et al., 1995; Nichols et al., 1994) indicate that all 22Ne in low-density graphite grains is from the decay of 22Na that was produced in the O/Ne zone in supernovae. One may argue why implanted 20,22Ne was not observed in the grains, considering the fact that the mass fraction of 20Ne is 5 orders of magnitude larger than that of 22Na. Croat et al. (2003) observed TiC subgrains inside low-density graphite grains have amorphous rims with the thickness of 3 to 15 nm, indicating atom bombardment from the surrounding gas. Assuming the gas is He, they estimated the velocity is 50 km/s or less. If the relative velocities between the Ne and the graphite grains are in that range, the penetration depth into the graphite grains is 2nm. Such shallow surface layers would be sputtered once the grains hit the reverse shock and keep traveling into the hot H-rich region (Nozawa et al, 2007). It remains to be seen whether or not 22Na in higher-density graphite is from supernovae or novae, or both. Amari, S. et al. 1995, Geochim. Cosmochim. Acta, 59, 1411 Croat, T.K. et al. 2003, Geochim. Cosmochim. Acta, 67, 4705 Nichols R.H. et al. 1994, Meteoritics, 29, 510 Nozawa, T. et al. 2007 ApJ, 666, 955

An Ultrasonic Scanning Technique for In-Situ `Bowing' Measurement of Prototype Fast Breeder Reactor Fuel Sub-Assembly

NASA Astrophysics Data System (ADS)

Swaminathan, K.; Asokane, C.; Sylvia, J. I.; Kalyanasundaram, P.; Swaminathan, P.

2012-02-01

An ultrasonic under-sodium scanner has been developed for deployment in Prototype Fast Breeder Reactor (PFBR) which is in advanced stage of construction at Kalpakkam, India. Its purpose is to scan the above-core plenum for detection, if any, of displacement of sub-assemblies. During its burn-up in the reactor, the head of a Fuel Sub-Assembly (FSA) may undergo a lateral shift from its original position (called `bowing') due to the fast neutron induced damage on its structural material. A simple scanning technique has been developed for measuring the extent of bowing in-situ. This paper describes a PC-controlled mock-up of the scanner used to implement the scanning technique and the results obtained of scanning a mock-up FSA head under water. The details of the liquid-sodium proof transducer developed for use in the PFBR scanner and its performance are also discussed.

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

DOE PAGES

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

2018-02-02

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

Dynamic event tree analysis with the SAS4A/SASSYS-1 safety analysis code

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

Jankovsky, Zachary K.; Denman, Matthew R.; Aldemir, Tunc

The consequences of a transient in an advanced sodium-cooled fast reactor are difficult to capture with the traditional approach to probabilistic risk assessment (PRA). Numerous safety-relevant systems are passive and may have operational states that cannot be represented by binary success or failure. In addition, the specific order and timing of events may be crucial which necessitates the use of dynamic PRA tools such as ADAPT. The modifications to the SAS4A/SASSYS-1 sodium-cooled fast reactor safety analysis code for linking it to ADAPT to perform a dynamic PRA are described. A test case is used to demonstrate the linking process andmore » to illustrate the type of insights that may be gained with this process. Finally, newly-developed dynamic importance measures are used to assess the significance of reactor parameters/constituents on calculated consequences of initiating events.« less

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL

DOEpatents

Barton, C.J.; Grimes, W.R.

1960-01-01

Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

Transient Effects in Turbulence Modelling.

DTIC Science & Technology

1979-12-01

plenum region of a liquid-metal- cooled fast breeder reactor (LMFBR). The efficient heat transfer characteristics of liquid metal coolant, combined...Transients in Generalized Liquid-Metal Fast Breeder Reactor Outlet Plenums," Nuclear Technology, Vol. 44, July 1979, p. 210. 135 15. Lorenz, J. J., "MIX... Sodium Coolant in the Outlet Plenum of a Fast Nuclear Reactor ," Int. J. Heat Mass Transfer, Vol. 21, 1978, pp. 1565-1579. 19. Chen, Y. B., Golay, M. W

A Nanoporous Carbon/Exfoliated Graphite Composite For Supercapacitor Electrodes

NASA Astrophysics Data System (ADS)

Rosi, Memoria; Ekaputra, Muhamad P.; Iskandar, Ferry; Abdullah, Mikrajuddin; Khairurrijal

2010-12-01

Nanoporous carbon was prepared from coconut shells using a simple heating method. The nanoporous carbon is subjected to different treatments: without activation, activation with polyethylene glycol (PEG), and activation with sodium hydroxide (NaOH)-PEG. The exfoliated graphite was synthesized from graphite powder oxidized with zinc acetate (ZnAc) and intercalated with polyvinyl alcohol (PVA) and NaOH. A composite was made by mixing the nanoporous carbon with NaOH-PEG activation, the exfoliated graphite and a binder of PVA solution, grinding the mixture, and annealing it using ultrasonic bath for 1 hour. All of as-synthesized materials were characterized by employing a scanning electron microscope (SEM), a MATLAB's image processing toolbox, and an x-ray diffractometer (XRD). It was confirmed that the composite is crystalline with (002) and (004) orientations. In addition, it was also found that the composite has a high surface area, a high distribution of pore sizes less than 40 nm, and a high porosity (67%). Noting that the pore sizes less than 20 nm are significant for ionic species storage and those in the range of 20 to 40 nm are very accessible for ionic clusters mobility across the pores, the composite is a promising material for the application as supercapacitor electrodes.

THETRIS: A MICRO-SCALE TEMPERATURE AND GAS RELEASE MODEL FOR TRISO FUEL

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

J. Ortensi; A.M. Ougouag

2011-12-01

The dominating mechanism in the passive safety of gas-cooled, graphite-moderated, high-temperature reactors (HTRs) is the Doppler feedback effect. These reactor designs are fueled with sub-millimeter sized kernels formed into TRISO particles that are imbedded in a graphite matrix. The best spatial and temporal representation of the feedback effect is obtained from an accurate approximation of the fuel temperature. Most accident scenarios in HTRs are characterized by large time constants and slow changes in the fuel and moderator temperature fields. In these situations a meso-scale, pebble and compact scale, solution provides a good approximation of the fuel temperature. Micro-scale models aremore » necessary in order to obtain accurate predictions in faster transients or when parameters internal to the TRISO are needed. Since these coated particles constitute one of the fundamental design barriers for the release of fission products, it becomes important to understand the transient behavior inside this containment system. An explicit TRISO fuel temperature model named THETRIS has been developed and incorporated into the CYNOD-THERMIX-KONVEK suite of coupled codes. The code includes gas release models that provide a simple predictive capability of the internal pressure during transients. The new model yields similar results to those obtained with other micro-scale fuel models, but with the added capability to analyze gas release, internal pressure buildup, and effects of a gap in the TRISO. The analyses show the instances when the micro-scale models improve the predictions of the fuel temperature and Doppler feedback. In addition, a sensitivity study of the potential effects on the transient behavior of high-temperature reactors due to the presence of a gap is included. Although the formation of a gap occurs under special conditions, its consequences on the dynamic behavior of the reactor can cause unexpected responses during fast transients. Nevertheless, the strong Doppler feedback forces the reactor to quickly stabilize.« less

Benchmark tests of JENDL-3.2 for thermal and fast reactors

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

Takano, Hideki; Akie, Hiroshi; Kikuchi, Yasuyuki

1994-12-31

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k{sub eff} and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k{sub eff} reactivity worths of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments.

Photodeposition of Gold, Platinum, or Silver onto Titanium Dioxide Nanoparticles at Steps of Highly Oriented Pyrolytic Graphite

NASA Astrophysics Data System (ADS)

Taing, James

The photodeposition of gold, platinum, or silver nanoparticles selectively onto isolated titanium dioxide (TiO2) nanoparticles created metal/TiO2 photocatalysts and heterogeneous catalysts, and validated the photocatalytic property of the semiconductor. The isolated and ordered TiO2 nanoparticles permitted clear observations of the stability, and changes in morphology, of the particles in various experimental conditions. The fabrication of TiO2 nanoparticles at the steps of highly oriented pyrolytic graphite (HOPG), utilizing physical vapor deposition, required heating the graphite substrate to a minimum of 800 °C. The production of a photocurrent, and plating of gold nanoparticles, confirmed the photocatalytic property of the TiO2 nanoparticles on HOPG when utilized as a photoelectrode in a two half-cell setup. Employing sodium chloride (1.0 M) as an electrolyte resulted in an increase/decrease of the photocurrent with the addition of gold cations to the half-cell without/with the TiO2 nanoparticles. A poor distribution of gold nanoparticles, roughly 40-45 nm wide, deposited around few of the TiO2 nanoparticles. A lower concentration of sodium chloride (0.1 M) resulted in a coalescence of Au nanoparticles, roughly 10 nm, around many TiO2 nanoparticles. Using sodium nitrate as an electrolyte resulted in a rapid decay in the photocurrent and a growth of an unidentified material on the TiO2 nanoparticles. The unidentified material hindered the reduction of gold cations introduced midway through the experiment. With gold cations present at the onset of the experiment, disperse gold nanoparticles (˜5-10 nm) deposited around the TiO2 nanoparticles. In the absence of additional electrolyte, many disperse gold nanoparticles less than 5 nm deposited onto the TiO2 nanoparticles. More platinum than gold selectively deposited onto the TiO2 nanoparticles. On the contrary, less silver selectively deposited onto the TiO2 nanoparticles. Scanning electron microscopy and atomic force microscopy determined the morphology and distribution of the TiO2 nanoparticles and metal/TiO 2 nanocomposites. Energy dispersive X-ray spectroscopy, transmission electron microscopy, and X-ray photoelectron spectroscopy identified the composition of the materials.

Code qualification of structural materials for AFCI advanced recycling reactors.

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

Natesan, K.; Li, M.; Majumdar, S.

2012-05-31

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Codemore » Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.« less

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

Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HXmore » channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.« less

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

Samin, Adib; Li, Xiang; Zhang, Jinsuo

For liquid-sodium-cooled fast nuclear reactor systems, it is crucial to understand the behavior of lanthanides and other potential fission products in liquid sodium or other liquid metal solutions such as liquid cesium-sodium. In this study, we focus on lanthanide behavior in liquid sodium. Using ab initio molecular dynamics, we found that the solubility of cerium in liquid sodium at 1000 K was less than 0.78 at. %, and the diffusion coefficient of cerium in liquid sodium was calculated to be 5.57 × 10{sup −9} m{sup 2}/s. Furthermore, it was found that cerium in small amounts may significantly alter the heat capacity of themore » liquid sodium system. Our results are consistent with the experimental results for similar materials under similar conditions.« less

Status of Chronic Oxidation Studies of Graphite

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

Contescu, Cristian I.; Mee, Robert W.

Graphite will undergo extremely slow, but continuous oxidation by traces of moisture that will be present, albeit at very low levels, in the helium coolant of HTGR. This chronic oxidation may cause degradation of mechanical strength and thermal properties of graphite components if a porous oxidation layer penetrates deep enough in the bulk of graphite components during the lifetime of the reactor. The current research on graphite chronic oxidation is motivated by the acute need to understand the behavior of each graphite grade during prolonged exposure to high temperature chemical attack by moisture. The goal is to provide the elementsmore » needed to develop predictive models for long-time oxidation behavior of graphite components in the cooling helium of HTGR. The tasks derived from this goal are: (1) Oxidation rate measurements in order to determine and validate a comprehensive kinetic model suitable for prediction of intrinsic oxidation rates as a function of temperature and oxidant gas composition; (2) Characterization of effective diffusivity of water vapor in the graphite pore system in order to account for the in-pore transport of moisture; and (3) Development and validation of a predictive model for the penetration depth of the oxidized layer, in order to assess the risk of oxidation caused damage of particular graphite grades after prolonged exposure to the environment of helium coolant in HTGR. The most important and most time consuming of these tasks is the measurement of oxidation rates in accelerated oxidation tests (but still under kinetic control) and the development of a reliable kinetic model. This report summarizes the status of chronic oxidation studies on graphite, and then focuses on model development activities, progress of kinetic measurements, validation of results, and improvement of the kinetic models. Analysis of current and past results obtained with three grades of showed that the classical Langmuir-Hinshelwood model cannot reproduce all data collected so far. Starting from here we propose a modification of the LH model to include temperature activation of graphite surface as a Boltzmann activation function. The enhanced Boltzmann-Langmuir-Hinshelwood model (BLH) was tested successfully on three grades of graphite. The model is a robust, comprehensive mathematical function that allows better fitting of experimental results spanning a wide range of temperature and partial pressures of water vapor and hydrogen. However, the model did not fit satisfactorily the data extracted from the old report on graphite H-451 oxidation by water.« less

Chemical Kinetics, Heat Transfer, and Sensor Dynamics Revisited in a Simple Experiment

ERIC Educational Resources Information Center

Sad, Maria E.; Sad, Mario R.; Castro, Alberto A.; Garetto, Teresita F.

2008-01-01

A simple experiment about thermal effects in chemical reactors is described, which can be used to illustrate chemical reactor models, the determination and validation of their parameters, and some simple principles of heat transfer and sensor dynamics. It is based in the exothermic reaction between aqueous solutions of sodium thiosulfate and…

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)

6. Workers laying up the graphite core of the 105B ...

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

6. Workers laying up the graphite core of the 105-B file. In the lower-left can be seen a portion of the rear face of the pile, the top of its shielding wall, and the gun barrels protruding through it. The inside of the front face of the pile and its gun barrels can be seen toward the upper-right side. The angled top of the front shielding wall can be seen in the picture. All four walls were "stepped" in this manner where they joined with another wall or the ceiling to form a "labyrinth" joint, so that radiation would not have a straight route through any gaps in the joints. D-3045 - B Reactor, Richland, Benton County, WA

Production of aluminum-silicon alloy and ferrosilicon and commercial purity aluminum by the direct reduction process. Third interim technical report, Phase C for the period 1980 July 1-1980 September 30

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

Bruno, M.J.

1980-10-01

Pilot reactor VSR-3 operation in the third quarter was directed to tapping molten alloy product. Modifications to the hearth region included a tapping furnace to maintain taphole temperature, a graphite ring filter to separate carbides from matal and an alumina liner to eliminate carbiding from reaction of alloy with the graphite hearth walls. Tapping was not successful, however, due to high alloy viscosity from a large concentration of carbides. Three runs were made on the pilot crystallizer to determine the effects of alloy composition, cooling rate, tamping rate, remelt temperature and rate on eutectic Al-Si yield.

Sodium-NaK engineering handbook. Volume III. Sodium systems, safety, handling, and instrumentation. [LMFBR

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

Foust, O J

1978-01-01

The handbook is intended for use by present and future designers in the Liquid Metals Fast Breeder Reactor (LMFBR) Program and by the engineering and scientific community performing other type investigation and exprimentation requiring high-temperature sodium and NaK technology. The arrangement of subject matter progresses from a technological discussion of sodium and sodium--potassium alloy (NaK) to discussions of varius categories and uses of hardware in sodium and NaK systems. Emphasis is placed on sodium and NaK as heat-transport media. Sufficient detail is included for basic understanding of sodium and NaK technology and of technical aspects of sodium and NaK componentsmore » and instrument systems. Information presented is considered adequate for use in feasibility studies and conceptual design, sizing components and systems, developing preliminary component and system descriptions, identifying technological limitations and problem areas, and defining basic constraints and parameters.« less

Power Conversion and Energy Storage System for a Fusion Reactor 3. Performance of Large Electric Power Equipment and Future View 3.1 Large Capacity Battery System -Sodium-Sulfur Battery-

NASA Astrophysics Data System (ADS)

Nakabayashi, Takashi

The Ford Motor Company proposed the principle of the sodium-sulfur battery based on a beta-alumina solid electrolyte in 1967. Accordingly, sodium-sulfur battery technology was initially developed primarily for electric vehicle applications. Later, the Tokyo Electric Power Company (TEPCO) selected the sodium-sulfur battery technology as the preferred system for a dispersed utility energy storage system to substitute for the pumped hydro energy storage system. NGK Insulators, Ltd. (NGK) and TEPCO have jointly carried out the development of the sodium-sulfur battery since 1984. In April 2002, TEPCO and NGK made the sodium-sulfur battery for use as an energy storage system commercially available.

Determination of indium in geological materials by electrothermal-atomization atomic absorption spectrometry with a tungsten-impregnated graphite furance

USGS Publications Warehouse

Zhou, L.; Chao, T.T.; Meier, A.L.

1984-01-01

The sample is fused with lithium metaborate and the melt is dissolved in 15% (v/v) hydrobromic acid. Iron(III) is reduced with ascorbic acid to avoid its coextraction with indium as the bromide into methyl isobutyl ketone. Impregnation of the graphite furnace with sodium tungstate, and the presence of lithium metaborate and ascorbic acid in the reaction medium improve the sensitivity and precision. The limits of determination are 0.025-16 mg kg-1 indium in the sample. For 22 geological reference samples containing more than 0.1 mg kg-1 indium, relative standard deviations ranged from 3.0 to 8.5% (average 5.7%). Recoveries of indium added to various samples ranged from 96.7 to 105.6% (average 100.2%). ?? 1984.

Composite Materials for Thermal Energy Storage: Enhancing Performance through Microstructures

PubMed Central

Ge, Zhiwei; Ye, Feng; Ding, Yulong

2014-01-01

Chemical incompatibility and low thermal conductivity issues of molten-salt-based thermal energy storage materials can be addressed by using microstructured composites. Using a eutectic mixture of lithium and sodium carbonates as molten salt, magnesium oxide as supporting material, and graphite as thermal conductivity enhancer, the microstructural development, chemical compatibility, thermal stability, thermal conductivity, and thermal energy storage performance of composite materials are investigated. The ceramic supporting material is essential for preventing salt leakage and hence provides a solution to the chemical incompatibility issue. The use of graphite gives a significant enhancement on the thermal conductivity of the composite. Analyses suggest that the experimentally observed microstructural development of the composite is associated with the wettability of the salt on the ceramic substrate and that on the thermal conduction enhancer. PMID:24591286

Sodium chloride-catalyzed oxidation of multiwalled carbon nanotubes for environmental benefit.

PubMed

Endo, Morinobu; Takeuchi, Kenji; Tajiri, Takeyuki; Park, Ki Chul; Wang, Feng; Kim, Yoong-Ahm; Hayashi, Takuya; Terrones, Mauricio; Dresselhaus, Mildred S

2006-06-22

A sodium chloride (NaCl) catalyst (0.1 w/w %) lowers the oxidation temperature of graphitized multiwalled carbon nanotubes: MWCNT-20 (diameter: 20-70 nm) and MWCNT-80 (diameter: 80-150 nm). The analysis of the reaction kinetics indicates that the oxidation of MWCNT-20 and MWCNT-80 mixed with no NaCl exhibits single reaction processes with activation energies of E(a) = 159 and 152 kJ mol(-1), respectively. The oxidation reaction in the presence of NaCl is shown to consist of two different reaction processes, that is, a first reaction and a second reaction process. The first reaction process is dominant at a low temperature of around 600 degrees C, while the second reaction process becomes more dominant than the first one in a higher temperature region. The activation energies of the first reaction processes (MWCNT-20: E(a1) = 35.7 kJ mol(-1); MWCNT-80: E(a1) = 43.5 kJ mol(-1)) are much smaller than those of the second reaction processes (MWCNT-20: E(a2) = 170 kJ mol(-1); MWCNT-80: E(a2) = 171 kJ mol(-1)). The comparison of the kinetic parameters and the results of the spectroscopic and microscopic analyses imply that the lowering of the oxidation temperature in the presence of NaCl results from the introduction of disorder into the graphitized MWCNTs (during the first reaction process), thus increasing the facility of the oxidation reaction of the disorder-induced nanotubes (in the second reaction process). It is found that the larger nanopits and cracks on the outer graphitic layers are caused by the catalytic effect of NaCl. Therefore, the NaCl-mixed samples showed more rapid and stronger oxidation compared with that of the nonmixed samples at the same residual quantity.

Preparation of RuO2-TiO2/Nano-graphite composite anode for electrochemical degradation of ceftriaxone sodium.

PubMed

Li, Dong; Guo, Xiaolei; Song, Haoran; Sun, Tianyi; Wan, Jiafeng

2018-06-05

Graphite-like material is widely used for preparing various electrodes for wastewater treatment. To enhance the electrochemical degradation efficiency of Nano-graphite (Nano-G) anode, RuO 2 -TiO 2 /Nano-G composite anode was prepared through the sol-gel method and hot-press technology. RuO 2 -TiO 2 /Nano-G composite was characterized by X-ray diffraction, X-ray photoelectron spectroscopy, transmission electron microscopy and N 2 adsorption-desorption. Results showed that RuO 2 , TiO 2 and Nano-G were composited successfully, and RuO 2 and TiO 2 nanoparticles were distributed uniformly on the surface of Nano-G sheet. Specific surface area of RuO 2 -TiO 2 /Nano-G composite was higher than that of TiO 2 /Nano-G composite and Nano-G. Electrochemical performances of RuO 2 -TiO 2 /Nano-G anode were investigated by cyclic voltammetry, electrochemical impedance spectroscopy. RuO 2 -TiO 2 /Nano-G anode was applied to electrochemical degradation of ceftriaxone. The generation of hydroxyl radical (OH) was measured. Results demonstrated that RuO 2 -TiO 2 /Nano-G anode displayed enhanced electrochemical degradation efficiency towards ceftriaxone and yield of OH, which is derived from the synergetic effect between RuO 2 , TiO 2 and Nano-G, which enhance the specific surface area, improve the electrochemical oxidation activity and lower the charge transfer resistance. Besides, the possible degradation intermediates and pathways of ceftriaxone sodium were identified. This study may provide a viable and promising prospect for RuO 2 -TiO 2 /Nano-G anode towards effective electrochemical degradation of antibiotics from wastewater. Copyright © 2018 Elsevier B.V. All rights reserved.

Surfactant exfoliated 2D hexagonal Boron Nitride (2D-hBN) explored as a potential electrochemical sensor for dopamine: surfactants significantly influence sensor capabilities.

PubMed

Khan, Aamar F; Brownson, Dale A C; Foster, Christopher W; Smith, Graham C; Banks, Craig E

2017-05-21

Surfactant exfoliated 2D hexagonal Boron Nitride (2D-hBN) nanosheets are explored as a potential electrochemical sensing platform and evaluated towards the electroanalytical sensing of dopamine (DA) in the presence of the common interferents, ascorbic acid (AA) and uric acid (UA). Surfactant exfoliated 2D-hBN nanosheets (2-4 layers) fabricated using sodium cholate in aqueous media are electrically wired via a drop-casting modification process onto disposable screen-printed graphite electrodes (SPEs). We critically evaluate the performance of these 2D-hBN modified SPEs and demonstrate the effect of 'mass coverage' towards the detection of DA, AA and UA. Previous studies utilising surfactant-free (pristine) 2D-hBN modified SPEs have shown a beneficial effect towards the detection of DA, AA and UA when compared to the underlying/unmodified graphite-based electrode. We show that the fabrication route utilised to prepare 2D-hBN is a vital experimental consideration, such that the beneficial effect previously reported is considerably reduced when surfactant exfoliated 2D-hBN is utilised. We demonstrate for the first time, through implementation of control experiments in the form of surfactant modified graphite electrodes, that sodium cholate is a major contributing factor to the aforementioned detrimental behaviour. The significance here is not in the material per se, but the fundamental knowledge of the surfactant and surface coverage changing the electrochemical properties of the material under investigation. Given the wide variety of ionic and non-ionic surfactants that are utilised in the manufacture of novel 2D materials, the control experiments reported herein need to be performed in order to de-convolute the electrochemical response and effectively evaluate the 'underlying surface/surfactant/2D materials' electrocatalytic contribution.

Heat dissipating nuclear reactor with metal liner

DOEpatents

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

1985-11-21

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

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

NASA Astrophysics Data System (ADS)

Blandinskiy, V. Yu.

2014-12-01

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

Heat dissipating nuclear reactor with metal liner

DOEpatents

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

1987-01-01

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

NEUTRONIC REACTOR

DOEpatents

Wigner, E.P.

1957-09-17

A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

Nuclear reactor control

DOEpatents

Cawley, William E.; Warnick, Robert F.

1982-01-01

1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

Revised FINAL–REPORT NO. 2: INDEPENDENT CONFIRMATORY SURVEY SUMMARY AND RESULTS FOR THE ENRICO FERMI ATOMIC POWER PLANT, UNIT 1, NEWPORT, MICHIGAN (DOCKET NO. 50 16; RFTA 10-004) 2018-SR-02-1

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

Erika Bailey

2011-10-27

The Enrico Fermi Atomic Power Plant, Unit 1 (Fermi 1) was a fast breeder reactor design that was cooled by sodium and operated at essentially atmospheric pressure. On May 10, 1963, the Atomic Energy Commission (AEC) granted an operating license, DPR-9, to the Power Reactor Development Company (PRDC), a consortium specifically formed to own and operate a nuclear reactor at the Fermi 1 site. The reactor was designed for a maximum capability of 430 megawatts (MW); however, the maximum reactor power with the first core loading (Core A) was 200 MW. The primary system was filled with sodium in Decembermore » 1960 and criticality was achieved in August 1963. The reactor was tested at low power during the first couple years of operation. Power ascension testing above 1 MW commenced in December 1965 immediately following the receipt of a high-power operating license. In October 1966 during power ascension, zirconium plates at the bottom of the reactor vessel became loose and blocked sodium coolant flow to some fuel subassemblies. Two subassemblies started to melt and the reactor was manually shut down. No abnormal releases to the environment occurred. Forty-two months later after the cause had been determined, cleanup completed, and the fuel replaced, Fermi 1 was restarted. However, in November 1972, PRDC made the decision to decommission Fermi 1 as the core was approaching its burn-up limit. The fuel and blanket subassemblies were shipped off-site in 1973. Following that, the secondary sodium system was drained and sent off-site. The radioactive primary sodium was stored on-site in storage tanks and 55 gallon (gal) drums until it was shipped off-site in 1984. The initial decommissioning of Fermi 1 was completed in 1975. Effective January 23, 1976, DPR-9 was transferred to the Detroit Edison Company (DTE) as a 'possession only' license (DTE 2010a). This report details the confirmatory activities performed during the second Oak Ridge Institute for Science and Education (ORISE) site visit to Fermi 1 in November 2010. The survey was strategically planned during a Unit 2 (Fermi 2) outage to take advantage of decreased radiation levels that were observed and attributed to Fermi 2 from the operating unit during the first site visit. However, during the second visit there were elevated radiation levels observed and attributed to the partially dismantled Fermi 1 reactor vessel and a waste storage box located on the 3rd floor of the Fermi 1 Turbine Building. Confirmatory surveys (unshielded) performed directly in the line of sight of these areas were affected. The objective of the confirmatory survey was to verify that the final radiological conditions were accurately and adequately described in Final Status Survey (FSS) documentation, relative to the established release criteria. This objective was achieved by performing document reviews, as well as independent measurements and sampling. Specifically, documentation of the planning, implementation, and results of the FSS were evaluated; side-by-side FSS measurement and source comparisons were performed; site areas were evaluated relative to appropriate FSS classification; and areas were assessed for residual, undocumented contamination.« less

Behavior of graphite under heat load and in contact with a hydrogen plasma

NASA Astrophysics Data System (ADS)

Bohdansky, J.; Croessmann, C. D.; Linke, J.; McDonald, J. M.; Morse, D. H.; Pontau, A. E.; Watson, R. D.; Whitley, J. B.; Goebel, D. M.; Hirooka, Y.; Leung, K.; Conn, R. W.; Roth, J.; Ottenberger, W.; Kotzlowski, H. E.

1987-05-01

Graphite is extensively used in large tokamaks today. In these machines the material is exposed to vacuum, to intense heat loads, and to the edge plasma. The use of graphite in such machines, therefore, depends on the outgassing behavior, the heat shock resistance, and thermochemical properties in a hydrogen plasma. Investigations of these properties made at different laboratories are described here. Experiments conducted at Sandia National Laboratories (SNL), Livermore, and the Max-Planck-Institut für Plasmaphysik (IPP) in Garching showed that the outgassing behavior of fine-grain reactor-grade graphite and carbon fiber composites depends on the pretreatment (manufacturing and/or storage). However, after proper outgassing the samples tested behave similarly in the case of fine-grain graphite, but the outgassing remains high for the carbon fiber composites. Heat shock tests have been made with the Electron Beam Test System (EBTS) at SNL, Albuquerque. Directly cooled graphite samples (FE 159 graphite brazed onto Mo tubes) showed no failure at a heat load of 700 W/cm 2, 20 s; or 10 kW, 1 s. Thermal erosion due to sublimination and particle emission from the graphite surface was observed. This effect is related to the surface temperature and becomes significant at temperatures above 2500°K. Fourteen different types of graphite were tested; the main differences among these samples were the different surface temperatures obtained under the same heating conditions. Cracking due to heat shocks was observed in some of the samples, but none of the carbon fiber composites failed. Thermochemical properties have been tested in the PISCES plasma generator at UCLA for ion energies of around 100 eV. The formation of C-H compounds was observed spectroscopically at sample temperatures of around 600°C. However, this chemical reaction did not lead to erosion as observed in beam experiments but to a drastic change of the surface structure due to redeposition. Carbon-hydrogen lines were still observed at sample temperatures of around 100°C. Under these conditions the erosion yield is high and in agreement with those measured in beam experiments.

Impact of nuclear data on sodium-cooled fast reactor calculations

NASA Astrophysics Data System (ADS)

Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

2016-03-01

Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors.

Apparatus for obtaining silicon from fluosilicic acid

DOEpatents

Sanjurjo, Angel

1986-05-20

Apparatus for producing low cost, high purity solar grade silicon ingots in single crystal or quasi single crystal ingot form in a substantially continuous operation in a two stage reactor starting with sodium fluosilicate and a metal more electropositive than silicon (preferably sodium) in separate compartments having easy vapor transport therebetween and thermally decomposing the sodium fluosilicate to cause formation of substantially pure silicon and a metal fluoride which may be continuously separated in the melt and silicon may be directly and continuously cast from the melt.

Electric field induced needle-pulsed arc discharge carbon nanotube production apparatus: circuitry and mechanical design.

PubMed

Kia, Kaveh Kazemi; Bonabi, Fahimeh

2012-12-01

A simple and low cost apparatus is reported to produce multiwall carbon nanotubes and carbon nano-onions by a low power short pulsed arc discharge reactor. The electric circuitry and the mechanical design details and a micro-filtering assembly are described. The pulsed-plasma is generated and applied between two graphite electrodes. The pulse width is 0.3 μs. A strong dc electric field is established along side the electrodes. The repetitive discharges occur in less than 1 mm distance between a sharp tip graphite rod as anode, and a tubular graphite as cathode. A hydrocarbon vapor, as carbon source, is introduced through the graphite nozzle in the cathode assembly. The pressure of the chamber is controlled by a vacuum pump. A magnetic field, perpendicular to the plasma path, is provided. The results show that the synergetic use of a pulsed-current and a dc power supply enables us to synthesize carbon nanoparticles with short pulsed plasma. The simplicity and inexpensiveness of this plan is noticeable. Pulsed nature of plasma provides some extra degrees of freedom that make the production more controllable. Effects of some design parameters such as electric field, pulse frequency, and cathode shape are discussed. The products are examined using scanning probe microscopy techniques.

Electric field induced needle-pulsed arc discharge carbon nanotube production apparatus: Circuitry and mechanical design

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

Kia, Kaveh Kazemi; Bonabi, Fahimeh

A simple and low cost apparatus is reported to produce multiwall carbon nanotubes and carbon nano-onions by a low power short pulsed arc discharge reactor. The electric circuitry and the mechanical design details and a micro-filtering assembly are described. The pulsed-plasma is generated and applied between two graphite electrodes. The pulse width is 0.3 {mu}s. A strong dc electric field is established along side the electrodes. The repetitive discharges occur in less than 1 mm distance between a sharp tip graphite rod as anode, and a tubular graphite as cathode. A hydrocarbon vapor, as carbon source, is introduced through themore » graphite nozzle in the cathode assembly. The pressure of the chamber is controlled by a vacuum pump. A magnetic field, perpendicular to the plasma path, is provided. The results show that the synergetic use of a pulsed-current and a dc power supply enables us to synthesize carbon nanoparticles with short pulsed plasma. The simplicity and inexpensiveness of this plan is noticeable. Pulsed nature of plasma provides some extra degrees of freedom that make the production more controllable. Effects of some design parameters such as electric field, pulse frequency, and cathode shape are discussed. The products are examined using scanning probe microscopy techniques.« less

Large-Scale Weibull Analysis of H-451 Nuclear- Grade Graphite Specimen Rupture Data

NASA Technical Reports Server (NTRS)

Nemeth, Noel N.; Walker, Andrew; Baker, Eric H.; Murthy, Pappu L.; Bratton, Robert L.

2012-01-01

A Weibull analysis was performed of the strength distribution and size effects for 2000 specimens of H-451 nuclear-grade graphite. The data, generated elsewhere, measured the tensile and four-point-flexure room-temperature rupture strength of specimens excised from a single extruded graphite log. Strength variation was compared with specimen location, size, and orientation relative to the parent body. In our study, data were progressively and extensively pooled into larger data sets to discriminate overall trends from local variations and to investigate the strength distribution. The CARES/Life and WeibPar codes were used to investigate issues regarding the size effect, Weibull parameter consistency, and nonlinear stress-strain response. Overall, the Weibull distribution described the behavior of the pooled data very well. However, the issue regarding the smaller-than-expected size effect remained. This exercise illustrated that a conservative approach using a two-parameter Weibull distribution is best for designing graphite components with low probability of failure for the in-core structures in the proposed Generation IV (Gen IV) high-temperature gas-cooled nuclear reactors. This exercise also demonstrated the continuing need to better understand the mechanisms driving stochastic strength response. Extensive appendixes are provided with this report to show all aspects of the rupture data and analytical results.

Electric field induced needle-pulsed arc discharge carbon nanotube production apparatus: Circuitry and mechanical design

NASA Astrophysics Data System (ADS)

Kia, Kaveh Kazemi; Bonabi, Fahimeh

2012-12-01

A simple and low cost apparatus is reported to produce multiwall carbon nanotubes and carbon nano-onions by a low power short pulsed arc discharge reactor. The electric circuitry and the mechanical design details and a micro-filtering assembly are described. The pulsed-plasma is generated and applied between two graphite electrodes. The pulse width is 0.3 μs. A strong dc electric field is established along side the electrodes. The repetitive discharges occur in less than 1 mm distance between a sharp tip graphite rod as anode, and a tubular graphite as cathode. A hydrocarbon vapor, as carbon source, is introduced through the graphite nozzle in the cathode assembly. The pressure of the chamber is controlled by a vacuum pump. A magnetic field, perpendicular to the plasma path, is provided. The results show that the synergetic use of a pulsed-current and a dc power supply enables us to synthesize carbon nanoparticles with short pulsed plasma. The simplicity and inexpensiveness of this plan is noticeable. Pulsed nature of plasma provides some extra degrees of freedom that make the production more controllable. Effects of some design parameters such as electric field, pulse frequency, and cathode shape are discussed. The products are examined using scanning probe microscopy techniques.

Crystallization and precipitation of phosphate from swine wastewater by magnesium metal corrosion.

PubMed

Huang, Haiming; Liu, Jiahui; Jiang, Yang

2015-11-12

This paper presents a unique approach for magnesium dosage in struvite precipitation by Mg metal corrosion. The experimental results showed that using an air bubbling column filled with Mg metal and graphite pellets for the magnesium dosage was the optimal operation mode, which could significantly accelerate the corrosion of the Mg metal pellets due to the presence of graphite granules. The reaction mechanism experiments revealed that the solution pH could be used as the indicator for struvite crystallization by the process. Increases in the Mg metal dosage, mass ratio of graphite and magnesium metal (G:M) and airflow rate could rapidly increase the solution pH. When all three conditions were at 10 g L(-1), 1:1 and 1 L min(-1), respectively, the phosphate recovery efficiency reached 97.5%. To achieve a high level of automation for the phosphate recovery process, a continuous-flow reactor immersed with the graphite-magnesium air bubbling column was designed to harvest the phosphate from actual swine wastewater. Under conditions of intermittently supplementing small amounts of Mg metal pellets, approximately 95% of the phosphate could be stably recovered as struvite of 95.8% (±0.5) purity. An economic analysis indicated that the process proposed was technically simple and economically feasible.

Crystallization and precipitation of phosphate from swine wastewater by magnesium metal corrosion

PubMed Central

Huang, Haiming; Liu, Jiahui; Jiang, Yang

2015-01-01

This paper presents a unique approach for magnesium dosage in struvite precipitation by Mg metal corrosion. The experimental results showed that using an air bubbling column filled with Mg metal and graphite pellets for the magnesium dosage was the optimal operation mode, which could significantly accelerate the corrosion of the Mg metal pellets due to the presence of graphite granules. The reaction mechanism experiments revealed that the solution pH could be used as the indicator for struvite crystallization by the process. Increases in the Mg metal dosage, mass ratio of graphite and magnesium metal (G:M) and airflow rate could rapidly increase the solution pH. When all three conditions were at 10 g L–1, 1:1 and 1 L min–1, respectively, the phosphate recovery efficiency reached 97.5%. To achieve a high level of automation for the phosphate recovery process, a continuous-flow reactor immersed with the graphite-magnesium air bubbling column was designed to harvest the phosphate from actual swine wastewater. Under conditions of intermittently supplementing small amounts of Mg metal pellets, approximately 95% of the phosphate could be stably recovered as struvite of 95.8% (±0.5) purity. An economic analysis indicated that the process proposed was technically simple and economically feasible. PMID:26558521

Graphite Microstructural Characterization Using Time-Domain and Correlation-Based Ultrasonics

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

Spicer, James

Among techniques that have been used to determine elastic modulus in nuclear graphites, ultrasonic methods have enjoyed wide use and standards using contacting piezoelectric tranducers have been developed to ensure repeatability of these types of measurements. However, the use of couplants and the pressures used to effectively couple transducers to samples can bias measurements and produce results that are not wholly related to the properties of the graphite itself. In this work, we have investigated the use of laser ultrasonic methods for making elastic modulus measurements in nuclear graphites. These methods use laser-based transmitters and receivers to gather data andmore » do not require use of ultrasonic couplants or mechanical contact with the sample. As a result, information directly related to the elastic responses of graphite can be gathered even if the graphite is porous, brittle and compliant. In particular, we have demonstrated the use of laser ultrasonics for the determination of both Young’s modulus and shear modulus in a range of nuclear graphites including those that are being considered for use in future nuclear reactors. These results have been analyzed to assess the contributions of porosity and microcracking to the elastic responses of these graphites. Laser-based methods have also been used to assess the moduli of NBG-18 and IG-110 where samples of each grade were oxidized to produce specific changes in porosity. These data were used to develop new models for the elastic responses of nuclear graphites and these models have been used to infer specific changes in graphite microstructure that occur during oxidation that affect elastic modulus. Specifically, we show how ultrasonic measurements in oxidized graphites are consistent with nano/microscale oxidation processes where basal plane edges react more readily than basal plane surfaces. We have also shown the use of laser-based methods to perform shear-wave birefringence measurements and have shown how these measurements can be used to assess elastic anisotropy in nuclear graphites. Using models developed in this program, ultrasonic data were interpreted to extract orientation distribution coefficients that could be used to represent anisotropy in these materials. This demonstration showed the use of ultrasonic methods to quantify anisotropy and how these methods provide more detailed information than do measurements of thermal expansion – a technique commonly used for assessing anisotropy in nuclear graphites. Finally, we have employed laser-based, ultrasonic-correlation techniques in attempts to quantify aspects of graphite microstructure such as pore size and distribution. Results of these measurements indicate that additional work must be performed to make this ultrasonic approach viable for quantitative microstructural characterization.« less

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

NASA Astrophysics Data System (ADS)

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

2000-05-01

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

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

DOE PAGES

Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; ...

2015-07-17

Building new fission reactors in the United States presents many technical and regulatory challenges. Chief among the technical challenges is the need to share and present results from new high- fidelity, high- performance simulations in an easily consumable way. In light of the modern multi-scale, multi-physics simulations can generate petabytes of data, this will require the development of new techniques and methods to reduce the data to familiar quantities of interest with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately availablemore » in the community and need to be developed. Our paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It enables easy qualitative and quantitative comparisons between simulation results with a graphical user interface and cross-platform, multi-language input- output libraries for use by developers to work with the data. One example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented along with a detailed discussion of the system s requirements and design.« less

A Study of Intumescent Reaction Mechanisms.

DTIC Science & Technology

1984-08-01

Neoprene Fibers Considered Asbestos Metal Fiber (Steel Wool) Kevlar Mica Glass Fiber Mineral Wool Graphite Refrasil -7- • . . . ’ .. I , "’ - . NADC-84 170...Kevlar o Flexible Epoxy e Metal Fiber (Steel Wool) * Mineral Wool Fillers * Borax * Sodium Metasilicate * Ammonium Phosphate o Aluminum Sulfate...reference tables in Section 2.0] . The mineral wool appears to be the least effec- tive of the five fibers. To optimize thermal performance for a set of

The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

NASA Astrophysics Data System (ADS)

Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

2016-02-01

The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

ATOMIC POWER PLANT

DOEpatents

Daniels, F.

1957-11-01

This patent relates to neutronic reactor power plants and discloses a design of a reactor utilizing a mixture of discrete units of a fissionable material, such as uranium carbide, a neutron moderator material, such as graphite, to carry out the chain reaction. A liquid metal, such as bismuth, is used as the coolant and is placed in the reactor chamber with the fissionable and moderator material so that it is boiled by the heat of the reaction, the boiling liquid and vapors passing up through the interstices between the discrete units. The vapor and flue gases coming off the top of the chamber are passed through heat exchangers, to produce steam, for example, and thence through condensers, the condensed coolant being returned to the chamber by gravity and the non- condensible gases being carried off through a stack at the top of the structure.

The early development of neutron diffraction: Science in the wings of the Manhattan Project

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

Mason, Thom; Gawne, Timothy J; Nagler, Stephen E

2012-01-01

Although neutron diffraction was first observed using radioactive decay sources shortly after the discovery of the neutron, it was only with the availability of higher intensity neutron beams from the first nuclear reactors, constructed as part of the Manhattan project, that systematic investigation of Bragg scattering became possible. Remarkably, at a time when the war effort was singularly focused on the development of the atomic bomb, groups working at Oak Ridge and Chicago carried out key measurements and recognized the future utility of neutron diffraction quite independent of its contributions to the measurements of nuclear cross sections. Ernest O. Wollan,more » Lyle B. Borst, and Walter H. Zinn were all able to observe neutron diffraction in 1944 using the X-10 graphite reactor and the CP-3 heavy water reactor.« less

Design of a single chambered microbial electrolytic cell reactor for production of biohydrogen from rice straw hydrolysate.

PubMed

Gupta, Pratima; Parkhey, Piyush

2015-06-01

Rice straw was pretreated using a microwave-assisted alkali pretreatment method. Cellulose recovery was approximately 82 %. This material was hydrolysed in an optimized enzymatic saccharification reaction using cellulase from Lysinibacillus sphaericus. This resulted in saccharification of 49 % of cellulosic biomass into glucose. A single chambered microbial electrolytic cell reactor of volume 2l was built using acrylic plastic sheets with graphite sheet as anode and a stainless-steel mesh as cathode. Shewanella putrefaciens was used as exoelectrogen to oxidize rice straw hydrolysate in the reactor for electrohydrogenesis. The maximum H2 yield obtained was 801 ml H2 g(-1) COD removal. Coulombic efficiency of 88 %, cathodic H2 recovery of 58 % and total H2 recovery of 51 % with an energy efficiency of 74 % were recorded.

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2010 CFR

2010-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2012 CFR

2012-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2011 CFR

2011-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2013 CFR

2013-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

10 CFR 50.62 - Requirements for reduction of risk from anticipated transients without scram (ATWS) events for...

Code of Federal Regulations, 2014 CFR

2014-01-01

... into the reactor pressure vessel a borated water solution at such a flow rate, level of boron concentration and boron-10 isotope enrichment, and accounting for reactor pressure vessel volume, that the... minute of 13 weight percent sodium pentaborate decahydrate solution at the natural boron-10 isotope...

Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor

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

Sham, Sam; Tan, Lizhen; Yamamoto, Yukinori

2013-01-01

Ferritic-martensitic (FM) steel Grade 92, with or without thermomechanical treatment (TMT), and austenitic stainless steels HT-UPS (high-temperature ultrafine precipitate strengthening) and NF709 were selected as potential candidate structural materials in the U.S. Sodium-cooled Fast Reactor (SFR) program. The objective is to develop advanced steels with improved properties as compared with reference materials such as Grade 91 and Type 316H steels that are currently in nuclear design codes. Composition modification and/or processing optimization (e.g., TMT and cold-work) were performed to improve properties such as resistance to thermal aging, creep, creep-fatigue, fracture, and sodium corrosion. Testings to characterize these properties for themore » advanced steels were conducted by the Idaho National Laboratory, the Argonne National Laboratory and the Oak Ridge National Laboratory under the U.S. SFR program. This paper focuses on the resistance to thermal aging and creep of the advanced steels. The advanced steels exhibited up to two orders of magnitude increase in creep life compared to the reference materials. Preliminary results on the weldment performance of the advanced steels are also presented. The superior performance of the advanced steels would improve reactor design flexibility, safety margins and economics.« less

Apparatus for detecting leakage of liquid sodium

DOEpatents

Himeno, Yoshiaki

1978-01-01

An apparatus for detecting the leakage of liquid sodium includes a cable-like sensor adapted to be secured to a wall of piping or other equipment having sodium on the opposite side of the wall, and the sensor includes a core wire electrically connected to the wall through a leak current detector and a power source. An accidental leakage of the liquid sodium causes the corrosion of a metallic layer and an insulative layer of the sensor by products resulted from a reaction of sodium with water or oxygen in the atmospheric air so as to decrease the resistance between the core wire and the wall. Thus, the leakage is detected as an increase in the leaking electrical current. The apparatus is especially adapted for use in detecting the leakage of liquid sodium from sodium-conveying pipes or equipment in a fast breeder reactor.

Development and applications of methodologies for the neutronic design of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR)

NASA Astrophysics Data System (ADS)

Fratoni, Massimiliano

This study investigated the neutronic characteristics of the Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a novel nuclear reactor concept that combines liquid salt (7LiF-BeF2---flibe) cooling and TRISO coated-particle fuel technology. The use of flibe enables operation at high power density and atmospheric pressure and improves passive decay-heat removal capabilities, but flibe, unlike conventional helium coolant, is not transparent to neutrons. The flibe occupies 40% of the PB-AHTR core volume and absorbs ˜8% of the neutrons, but also acts as an effective neutron moderator. Two novel methodologies were developed for calculating the time dependent and equilibrium core composition: (1) a simplified single pebble model that is relatively fast; (2) a full 3D core model that is accurate and flexible but computationally intensive. A parametric analysis was performed spanning a wide range of fuel kernel diameters and graphite-to-heavy metal atom ratios to determine the attainable burnup and reactivity coefficients. Using 10% enriched uranium ˜130 GWd/tHM burnup was found to be attainable, when the graphite-to-heavy metal atom ratio (C/HM) is in the range of 300 to 400. At this or smaller C/HM ratio all reactivity coefficients examined---coolant temperature, coolant small and full void, fuel temperature, and moderator temperature, were found to be negative. The PB-AHTR performance was compared to that of alternative options for HTRs, including the helium-cooled pebble-bed reactor and prismatic fuel reactors, both gas-cooled and flibe-cooled. The attainable burnup of all designs was found to be similar. The PB-AHTR generates at least 30% more energy per pebble than the He-cooled pebble-bed reactor. Compared to LWRs the PB-AHTR requires 30% less natural uranium and 20% less separative work per unit of electricity generated. For deep burn TRU fuel made from recycled LWR spent fuel, it was found that in a single pass through the core ˜66% of the TRU can be transmuted; this burnup is slightly superior to that attainable in helium-cooled reactors. A preliminary analysis of the modular variant for the PB-AHTR investigated the triple heterogeneity of this design and determined its performance characteristics.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas M.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Nuclear Thermal Rocket Simulation in NPSS

NASA Technical Reports Server (NTRS)

Belair, Michael L.; Sarmiento, Charles J.; Lavelle, Thomas L.

2013-01-01

Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.

Mars Mission Analysis Trades Based on Legacy and Future Nuclear Propulsion Options

NASA Astrophysics Data System (ADS)

Joyner, Russell; Lentati, Andrea; Cichon, Jaclyn

2007-01-01

The purpose of this paper is to discuss the results of mission-based system trades when using a nuclear thermal propulsion (NTP) system for Solar System exploration. The results are based on comparing reactor designs that use a ceramic-metallic (CERMET), graphite matrix, graphite composite matrix, or carbide matrix fuel element designs. The composite graphite matrix and CERMET designs have been examined for providing power as well as propulsion. Approaches to the design of the NTP to be discussed will include an examination of graphite, composite, carbide, and CERMET core designs and the attributes of each in regards to performance and power generation capability. The focus is on NTP approaches based on tested fuel materials within a prismatic fuel form per the Argonne National Laboratory testing and the ROVER/NERVA program. NTP concepts have been examined for several years at Pratt & Whitney Rocketdyne for use as the primary propulsion for human missions beyond earth. Recently, an approach was taken to examine the design trades between specific NTP concepts; NERVA-based (UC)C-Graphite, (UC,ZrC)C-Composite, (U,Zr)C-Solid Carbide and UO2-W CERMET. Using Pratt & Whitney Rocketdyne's multidisciplinary design analysis capability, a detailed mission and vehicle model has been used to examine how several of these NTP designs impact a human Mars mission. Trends for the propulsion system mass as a function of power level (i.e. thrust size) for the graphite-carbide and CERMET designs were established and correlated against data created over the past forty years. These were used for the mission trade study. The resulting mission trades presented in this paper used a comprehensive modeling approach that captures the mission, vehicle subsystems, and NTP sizing.

Silicon halide-alkali metal flames as a source of solar grade silicon

NASA Technical Reports Server (NTRS)

Olsen, D. B.; Miller, W. J.

1979-01-01

The feasibility of using alkali metal-silicon halide diffusion flames to produce solar-grade silicon in large quantities and at low cost is demonstrated. Prior work shows that these flames are stable and that relatively high purity silicon can be produced using Na + SiCl4 flames. Silicon of similar purity is obtained from Na + SiF4 flames although yields are lower and product separation and collection are less thermochemically favored. Continuous separation of silicon from the byproduct alkali salt was demonstrated in a heated graphite reactor. The process was scaled up to reduce heat losses and to produce larger samples of silicon. Reagent delivery systems, scaled by a factor of 25, were built and operated at a production rate of 0.5 kg Si/h. Very rapid reactor heating rates are observed with wall temperatures reaching greater than 2000 K. Heat release parameters were measured using a cooled stainless steel reactor tube. A new reactor was designed.

Addition of acetate improves stability of power generation using microbial fuel cells treating domestic wastewater.

PubMed

Stager, Jennifer L; Zhang, Xiaoyuan; Logan, Bruce E

2017-12-01

Power generation using microbial fuel cells (MFCs) must provide stable, continuous conversion of organic matter in wastewaters into electricity. However, when relatively small diameter (0.8cm) graphite fiber brush anodes were placed close to the cathodes in MFCs, power generation was unstable during treatment of low strength domestic wastewater. One reactor produced 149mW/m 2 before power generation failed, while the other reactor produced 257mW/m 2 , with both reactors exhibiting severe power overshoot in polarization tests. Using separators or activated carbon cathodes did not result in stable operation as the reactors continued to exhibit power overshoot based on polarization tests. However, adding acetate (1g/L) to the wastewater produced stable performance during fed batch and continuous flow operation, and there was no power overshoot in polarization tests. These results highlight the importance of wastewater strength and brush anode size for producing stable and continuous power in compact MFCs. Copyright © 2017 Elsevier B.V. All rights reserved.

CRBR pump water test experience

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

Cook, M.E.; Huber, K.A.

1983-01-01

The hydraulic design features and water testing of the hydraulic scale model and prototype pump of the sodium pumps used in the primary and intermediate sodium loops of the Clinch River Breeder Reactor Plant (CRBRP) are described. The Hydraulic Scale Model tests are performed and the results of these tests are discussed. The Prototype Pump tests are performed and the results of these tests are discussed.

Fast reactor safety and related physics. Volume IV. Phenomenology

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

Not Available

1976-01-01

Separate abstracts are included for 58 papers concerning single-phase flow and sodium boiling; sodium boiling and subassembly flow blockages; transient-overpower and loss-of-flow experiments; fuel and cladding behavior and relocation; fuel and cladding freezing; molten-fuel-coolant interaction; aerosols and fission product release, and post-accident heat removal. Thirteen papers have been perivously abstracted and included in ERA.

Neutron Fluence And DPA Rate Analysis In Pebble-Bed HTR Reactor Vessel Using MCNP

NASA Astrophysics Data System (ADS)

Hamzah, Amir; Suwoto; Rohanda, Anis; Adrial, Hery; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

In the Pebble-bed HTR reactor, the distance between the core and the reactor vessel is very close and the media inside are carbon and He gas. Neutron moderation capability of graphite material is theoretically lower than that of water-moderated reactors. Thus, it is estimated much more the fast neutrons will reach the reactor vessel. The fast neutron collisions with the atoms in the reactor vessel will result in radiation damage and could be reducing the vessel life. The purpose of this study was to obtain the magnitude of neutron fluence in the Pebble-bed HTR reactor vessel. Neutron fluence calculations in the pebble-bed HTR reactor vessel were performed using the MCNP computer program. By determining the tally position, it can be calculated flux, spectrum and neutron fluence in the position of Pebble-bed HTR reactor vessel. The calculations results of total neutron flux and fast neutron flux in the reactor vessel of 1.82x108 n/cm2/s and 1.79x108 n/cm2/s respectively. The fast neutron fluence in the reactor vessel is 3.4x1017 n/cm2 for 60 years reactor operation. Radiation damage in stainless steel material caused by high-energy neutrons (> 1.0 MeV) will occur when it has reached the neutron flux level of 1.0x1024 n/cm2. The neutron fluence results show that there is no radiation damage in the Pebble-bed HTR reactor vessel, so it is predicted that it will be safe to operate at least for 60 years.

Preliminary CFD study of Pebble Size and its Effect on Heat Transfer in a Pebble Bed Reactor

NASA Astrophysics Data System (ADS)

Jones, Andrew; Enriquez, Christian; Spangler, Julian; Yee, Tein; Park, Jungkyu; Farfan, Eduardo

2017-11-01

In pebble bed reactors, the typical pebble diameter used is 6cm, and within each pebble is are thousands of nuclear fuel kernels. However, efficiency of the reactor does not solely depend on the number of kernels of fuel within each graphite sphere, but also depends on the type and motion of the coolant within the voids between the spheres and the reactor itself. In this work a physical analysis of the pebble bed nuclear reactor's fluid dynamics is undertaken using Computational Fluid Dynamics software. The primary goal of this work is to observe the relationship between the different pebble diameters in an idealized alignment and the thermal transport efficiency of the reactor. The model constructed of our idealized argument will consist on stacked 8 pebble columns that fixed at the inlet on the reactor. Two different pebble sizes 4 cm and 6 cm will be studied and helium will be supplied as coolant with a fixed flow rate of 96 kg/s, also a fixed pebble surface temperatures will be used. Comparison will then be made to evaluate the efficiency of coolant to transport heat due to the varying sizes of the pebbles. Assistant Professor for the Department of Civil and Construction Engineering PhD.

Submersible sodium pump

DOEpatents

Brynsvold, Glen V.; Lopez, John T.; Olich, Eugene E.; West, Calvin W.

1989-01-01

An electromagnetic submerged pump has an outer cylindrical stator with an inner cylindrical conductive core for the submerged pumping of sodium in the cylindrical interstitial volume defined between the stator and core. The cylindrical interstitial volume is typically vertically oriented, and defines an inlet at the bottom and an outlet at the top. The outer stator generates upwardly conveyed toroidal magnetic fields, which fields convey preferably from the bottom of the pump to the top of the pump liquid sodium in the cold leg of a sodium cooled nuclear reactor. The outer cylindrical stator has a vertically disposed duct surrounded by alternately stacked layers of coil units and laminates.

Submersible sodium pump

DOEpatents

Brynsvold, G.V.; Lopez, J.T.; Olich, E.E.; West, C.W.

1989-11-21

An electromagnetic submerged pump has an outer cylindrical stator with an inner cylindrical conductive core for the submerged pumping of sodium in the cylindrical interstitial volume defined between the stator and core. The cylindrical interstitial volume is typically vertically oriented, and defines an inlet at the bottom and an outlet at the top. The outer stator generates upwardly conveyed toroidal magnetic fields, which fields convey preferably from the bottom of the pump to the top of the pump liquid sodium in the cold leg of a sodium cooled nuclear reactor. The outer cylindrical stator has a vertically disposed duct surrounded by alternately stacked layers of coil units and laminates. 14 figs.

Neutron Reference Benchmark Field Specifications: ACRR Polyethylene-Lead-Graphite (PLG) Bucket Environment (ACRR-PLG-CC-32-CL).

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

Vega, Richard Manuel; Parm, Edward J.; Griffin, Patrick J.

2015-07-01

This report was put together to support the International Atomic Energy Agency (IAEA) REAL- 2016 activity to validate the dosimetry community’s ability to use a consistent set of activation data and to derive consistent spectral characterizations. The report captures details of integral measurements taken in the Annular Core Research Reactor (ACRR) central cavity with the Polyethylene-Lead-Graphite (PLG) bucket, reference neutron benchmark field. The field is described and an “a priori” calculated neutron spectrum is reported, based on MCNP6 calculations, and a subject matter expert (SME) based covariance matrix is given for this “a priori” spectrum. The results of 37 integralmore » dosimetry measurements in the neutron field are reported.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

DOE PAGES

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel--coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

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

Sofu, Tanju

2015-04-01

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperaturemore » profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain cool-able. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.« less

Determination of Neutron Spectra in a Graphite Sphere for Fusion Reactor Studies

NASA Astrophysics Data System (ADS)

Bashter, I. B.; Cooper, P. N.

Calculated and experimental results for the neutron spectra at different radii in a graphite sphere irradiated with 14.1 MeV neutrons were shown to be in satisfactory agreement over the energy range 14.1 to 1.8 MeV neutrons. A group of curves were constructed which gives the radius of a graphite sphere shield required to attenuate the neutron intensity to a certain value. The data set used in the present work, with carbon-12 cross section, is shown to be useful for spherical calculations.Translated AbstractDie Bestimmung der Neutronenspektren in einer GraphitkugelDie Übereinstimmung experimentell bestimmter und berechneter Neutronenspektren in Abhängigkeit vom Ort in einer Graphitkugel wird in einem Energiebereich von 14,1 bis 1,8 MeV (bei einer Ausgangsenergie von 14,1 MeV je Neutron) gezeigt. Eine Gruppe von Kurven wird konstruiert, die den für eine bestimmte Dämpfung der Neutronenintensität notwendigen Radius einer Graphitkugel angeben. Es wird nachgewiesen, daß die in der Arbeit benutzte Datenbank für den 12C-Wirkungsquerschnitt in sphärischen Geometrien anwendbar ist.

Composite materials for thermal energy storage: enhancing performance through microstructures.

PubMed

Ge, Zhiwei; Ye, Feng; Ding, Yulong

2014-05-01

Chemical incompatibility and low thermal conductivity issues of molten-salt-based thermal energy storage materials can be addressed by using microstructured composites. Using a eutectic mixture of lithium and sodium carbonates as molten salt, magnesium oxide as supporting material, and graphite as thermal conductivity enhancer, the microstructural development, chemical compatibility, thermal stability, thermal conductivity, and thermal energy storage performance of composite materials are investigated. The ceramic supporting material is essential for preventing salt leakage and hence provides a solution to the chemical incompatibility issue. The use of graphite gives a significant enhancement on the thermal conductivity of the composite. Analyses suggest that the experimentally observed microstructural development of the composite is associated with the wettability of the salt on the ceramic substrate and that on the thermal conduction enhancer. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Capacitance‐Assisted Sustainable Electrochemical Carbon Dioxide Mineralisation

PubMed Central

Lamb, Katie J.; Dowsett, Mark R.; Chatzipanagis, Konstantinos; Scullion, Zhan Wei; Kröger, Roland; Lee, James D.

2017-01-01

Abstract An electrochemical cell comprising a novel dual‐component graphite and Earth‐crust abundant metal anode, a hydrogen producing cathode and an aqueous sodium chloride electrolyte was constructed and used for carbon dioxide mineralisation. Under an atmosphere of 5 % carbon dioxide in nitrogen, the cell exhibited both capacitive and oxidative electrochemistry at the anode. The graphite acted as a supercapacitive reagent concentrator, pumping carbon dioxide into aqueous solution as hydrogen carbonate. Simultaneous oxidation of the anodic metal generated cations, which reacted with the hydrogen carbonate to give mineralised carbon dioxide. Whilst conventional electrochemical carbon dioxide reduction requires hydrogen, this cell generates hydrogen at the cathode. Carbon capture can be achieved in a highly sustainable manner using scrap metal within the anode, seawater as the electrolyte, an industrially relevant gas stream and a solar panel as an effective zero‐carbon energy source. PMID:29171724

Graphite nanocomposites sensor for multiplex detection of antioxidants in food.

PubMed

Ng, Khan Loon; Tan, Guan Huat; Khor, Sook Mei

2017-12-15

Butylated hydroxyanisole (BHA), butylated hydroxytoluene (BHT), and tert-butylhydroquinone (TBHQ) are synthetic antioxidants used in the food industry. Herein, we describe the development of a novel graphite nanocomposite-based electrochemical sensor for the multiplex detection and measurement of BHA, BHT, and TBHQ levels in complex food samples using a linear sweep voltammetry technique. Moreover, our newly established analytical method exhibited good sensitivity, limit of detection, limit of quantitation, and selectivity. The accuracy and reliability of analytical results were challenged by method validation and comparison with the results of the liquid chromatography method, where a linear correlation of more than 0.99 was achieved. The addition of sodium dodecyl sulfate as supporting additive further enhanced the LSV response (anodic peak current, I pa ) of BHA and BHT by 2- and 20-times, respectively. Copyright © 2017 Elsevier Ltd. All rights reserved.

Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

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

Ponciroli, R.; Passerini, S.; Vilim, R. B.

In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based onmore » the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.« less

Corrosion-induced microstructural developments in 316 stainless steel during exposure to molten Li2BeF4(FLiBe) salt

NASA Astrophysics Data System (ADS)

Zheng, Guiqiu; He, Lingfeng; Carpenter, David; Sridharan, Kumar

2016-12-01

The microstructural developments in the near-surface regions of AISI 316 stainless steel during exposure to molten Li2BeF4 (FLiBe) salt have been investigated with the goal of using this material for the construction of the fluoride salt-cooled high-temperature reactor (FHR), a leading nuclear reactor concept for the next generation nuclear plants (NGNP). Tests were conducted in molten FLiBe salt (melting point: 459 °C) at 700 °C in graphite crucibles and 316 stainless steel crucibles for exposure duration of up to 3000 h. Corrosion-induced microstructural changes in the near-surface regions of the samples were characterized using scanning electron microscopy (SEM) in conjunction with energy dispersive x-ray spectroscopy (EDS) and electron backscatter diffraction (EBSD), and scanning transmission electron microscopy (STEM) with EDS capabilities. Intergranular corrosion attack in the near-surface regions was observed with associated Cr depletion along the grain boundaries. High-angle grain boundaries (15-180°) were particularly prone to intergranular attack and Cr depletion. The depth of attack extended to the depths of 22 μm after 3000-h exposure for the samples tested in graphite crucible, while similar exposure in 316 stainless steel crucible led to the attack depths of only about 11 μm. Testing in graphite crucibles led to the formation of nanometer-scale Mo2C, Cr7C3 and Al4C3 particle phases in the near-surface regions of the material. The copious depletion of Cr in the near-surface regions induced a γ-martensite to α-ferrite phase (FeNix) transformation. Based on the microstructural analysis, a thermal diffusion controlled corrosion model was developed and experimentally validated for predicting long-term corrosion attack depth.

Flow duct for nuclear reactors

DOEpatents

Straalsund, Jerry L.

1978-01-01

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

A summary of sodium-cooled fast reactor development

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

Aoto, Kazumi; Dufour, Philippe; Hongyi, Yang

Much of the basic technology for the Sodium-cooled fast Reactor (SFR) has been established through long term development experience with former fast reactor programs, and is being confirmed by the Phénix end-of-life tests in France, the restart of Monju in Japan, the lifetime extension of BN-600 in Russia, and the startup of the China Experimental Fast Reactor in China. Planned startup in 2014 for new SFRs: BN-800 in Russia and PFBR in India, will further enhance the confirmation of the SFR basic technology. Nowadays, the SFR development has advanced to aiming at establishment of the Generation-IV system which is dedicatedmore » to sustainable energy generation and actinide management, and several advanced SFR concepts are under development such as PRISM, JSFR, ASTRID, PGSFR, BN-1200, and CFR-600. Generation-IV International Forum is an international collaboration framework where various R&D activities are progressing on design of system and component, safety and operation, advanced fuel, and actinide cycle for the Generation-IV SFR development, and will play a beneficial role of promoting them thorough providing an opportunity to share the past experience and the latest data of design and R&D among countries developing SFR.« less

Theoretical Estimate of Maximum Possible Nuclear Explosion

DOE R&D Accomplishments Database

Bethe, H. A.

1950-01-31

The maximum nuclear accident which could occur in a Na-cooled, Be moderated, Pu and power producing reactor is estimated theoretically. (T.R.H.) 2O82 Results of nuclear calculations for a variety of compositions of fast, heterogeneous, sodium-cooled, U-235-fueled, plutonium- and power-producing reactors are reported. Core compositions typical of plate-, pin-, or wire-type fuel elements and with uranium as metal, alloy, and oxide were considered. These compositions included atom ratios in the following range: U-23B to U-235 from 2 to 8; sodium to U-235 from 1.5 to 12; iron to U-235 from 5 to 18; and vanadium to U-235 from 11 to 33. Calculations were performed to determine the effect of lead and iron reflectors between the core and blanket. Both natural and depleted uranium were evaluated as the blanket fertile material. Reactors were compared on a basis of conversion ratio, specific power, and the product of both. The calculated results are in general agreement with the experimental results from fast reactor assemblies. An analysis of the effect of new cross-section values as they became available is included. (auth)

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

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

Michael A. Pope

2011-10-01

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

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

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

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

2010-09-01

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

Eddy Current Flow Measurements in the FFTF

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

Nielsen, Deborah L.; Polzin, David L.; Omberg, Ronald P.

2017-02-02

The Fast Flux Test Facility (FFTF) is the most recent liquid metal reactor (LMR) to be designed, constructed, and operated by the U.S. Department of Energy (DOE). The 400-MWt sodium-cooled, fast-neutron flux reactor plant was designed for irradiation testing of nuclear reactor fuels and materials for liquid metal fast breeder reactors. Following shut down of the Clinch River Breeder Reactor Plant (CRBRP) project in 1983, FFTF continued to play a key role in providing a test bed for demonstrating performance of advanced fuel designs and demonstrating operation, maintenance, and safety of advanced liquid metal reactors. The FFTF Program provides valuablemore » information for potential follow-on reactor projects in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor control and operations. This report provides HEDL-TC-1344, “ECFM Flow Measurements in the FFTF Using Phase-Sensitive Detectors”, March 1979.« less

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR

DOEpatents

Szilard, L.; Young, G.J.

1958-03-01

This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

Articulated limiter blade for a tokamak fusion reactor

DOEpatents

Doll, D.W.

1982-10-21

A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

Articulated limiter blade for a tokamak fusion reactor

DOEpatents

Doll, David W.

1985-01-01

A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

Fiber optic sensors for nuclear power plant applications

NASA Astrophysics Data System (ADS)

Kasinathan, Murugesan; Sosamma, Samuel; BabuRao, Chelamchala; Murali, Nagarajan; Jayakumar, Tammana

2012-05-01

Studies have been carried out for application of Raman Distributed Temperature Sensor (RDTS) in Nuclear Power Plants (NPP). The high temperature monitoring in sodium circuits of Fast Breeder Reactor (FBR) is important. It is demonstrated that RDTS can be usefully employed in monitoring sodium circuits and in tracking the percolating sodium in the surrounding insulation in case of any leak. Aluminum Conductor Steel Reinforced (ACSR) cable is commonly used as overhead power transmission cable in power grid. The suitability of RDTS for detecting defects in ACSR overhead power cable, is also demonstrated.

Process and apparatus for obtaining silicon from fluosilicic acid

DOEpatents

Sanjurjo, Angel

1988-06-28

Process and apparatus for producing low cost, high purity solar grade silicon ingots in single crystal or quasi single crystal ingot form in a substantially continuous operation in a two stage reactor starting with sodium fluosilicate and a metal more electropositive than silicon (preferably sodium) in separate compartments having easy vapor transport therebetween and thermally decomposing the sodium fluosilicate to cause formation of substantially pure silicon and a metal fluoride which may be continuously separated in the melt and silicon may be directly and continuously cast from the melt.

Microscale Heat Conduction Models and Doppler Feedback

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

Hawari, Ayman I.; Ougouag, Abderrafi

2015-01-22

The objective of this project is to establish an approach for providing the fundamental input that is needed to estimate the magnitude and time-dependence of the Doppler feedback mechanism in Very High Temperature reactors. This mechanism is the foremost contributor to the passive safety of gas-cooled, graphite-moderated high temperature reactors that use fuel based on Tristructural-Isotropic (TRISO) coated particles. Therefore, its correct prediction is essential to the conduct of safety analyses for these reactors. Since the effect is directly dependent on the actual temperature reached by the fuel during transients, the underlying phenomena of heat deposition, heat transfer and temperaturemore » rise must be correctly predicted. To achieve the above objective, this project will explore an approach that accounts for lattice effects as well as local temperature variations and the correct definition of temperature and related local effects.« less

Simulations of carbon sputtering in fusion reactor divertor plates

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

Marian, J; Zepeda-Ruiz, L A; Gilmer, G H

2005-10-03

The interaction of edge plasma with material surfaces raises key issues for the viability of the International Thermonuclear Reactor (ITER) and future fusion reactors, including heat-flux limits, net material erosion, and impurity production. After exposure of the graphite divertor plate to the plasma in a fusion device, an amorphous C/H layer forms. This layer contains 20-30 atomic percent D/T bonded to C. Subsequent D/T impingement on this layer produces a variety of hydrocarbons that are sputtered back into the sheath region. We present molecular dynamics (MD) simulations of D/T impacts on amorphous carbon layer as a function of ion energymore » and orientation, using the AIREBO potential. In particular, energies are varied between 10 and 150 eV to transition from chemical to physical sputtering. These results are used to quantify yield, hydrocarbon composition and eventual plasma contamination.« less

Initial Neutronics Analyses for HEU to LEU Fuel Conversion of the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory

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

Kontogeorgakos, D.; Derstine, K.; Wright, A.

2013-06-01

The purpose of the TREAT reactor is to generate large transient neutron pulses in test samples without over-heating the core to simulate fuel assembly accident conditions. The power transients in the present HEU core are inherently self-limiting such that the core prevents itself from overheating even in the event of a reactivity insertion accident. The objective of this study was to support the assessment of the feasibility of the TREAT core conversion based on the present reactor performance metrics and the technical specifications of the HEU core. The LEU fuel assembly studied had the same overall design, materials (UO 2more » particles finely dispersed in graphite) and impurities content as the HEU fuel assembly. The Monte Carlo N–Particle code (MCNP) and the point kinetics code TREKIN were used in the analyses.« less

Multi-scale approach to the modeling of fission gas discharge during hypothetical loss-of-flow accident in gen-IV sodium fast reactor

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

Behafarid, F.; Shaver, D. R.; Bolotnov, I. A.

The required technological and safety standards for future Gen IV Reactors can only be achieved if advanced simulation capabilities become available, which combine high performance computing with the necessary level of modeling detail and high accuracy of predictions. The purpose of this paper is to present new results of multi-scale three-dimensional (3D) simulations of the inter-related phenomena, which occur as a result of fuel element heat-up and cladding failure, including the injection of a jet of gaseous fission products into a partially blocked Sodium Fast Reactor (SFR) coolant channel, and gas/molten sodium transport along the coolant channels. The computational approachmore » to the analysis of the overall accident scenario is based on using two different inter-communicating computational multiphase fluid dynamics (CMFD) codes: a CFD code, PHASTA, and a RANS code, NPHASE-CMFD. Using the geometry and time history of cladding failure and the gas injection rate, direct numerical simulations (DNS), combined with the Level Set method, of two-phase turbulent flow have been performed by the PHASTA code. The model allows one to track the evolution of gas/liquid interfaces at a centimeter scale. The simulated phenomena include the formation and breakup of the jet of fission products injected into the liquid sodium coolant. The PHASTA outflow has been averaged over time to obtain mean phasic velocities and volumetric concentrations, as well as the liquid turbulent kinetic energy and turbulence dissipation rate, all of which have served as the input to the core-scale simulations using the NPHASE-CMFD code. A sliding window time averaging has been used to capture mean flow parameters for transient cases. The results presented in the paper include testing and validation of the proposed models, as well the predictions of fission-gas/liquid-sodium transport along a multi-rod fuel assembly of SFR during a partial loss-of-flow accident. (authors)« less

Component and Technology Development for Advanced Liquid Metal Reactors

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

Anderson, Mark

2017-01-30

The following report details the significant developments to Sodium Fast Reactor (SFR) technologies made throughout the course of this funding. This report will begin with an overview of the sodium loop and the improvements made over the course of this research to make it a more advanced and capable facility. These improvements have much to do with oxygen control and diagnostics. Thus a detailed report of advancements with respect to the cold trap, plugging meter, vanadium equilibration loop, and electrochemical oxygen sensor is included. Further analysis of the university’s moving magnet pump was performed and included in a section ofmore » this report. A continuous electrical resistance based level sensor was built and tested in the sodium with favorable results. Materials testing was done on diffusion bonded samples of metal and the results are presented here as well. A significant portion of this work went into the development of optical fiber temperature sensors which could be deployed in an SFR environment. Thus, a section of this report presents the work done to develop an encapsulation method for these fibers inside of a stainless steel capillary tube. High temperature testing was then done on the optical fiber ex situ in a furnace. Thermal response time was also explored with the optical fiber temperature sensors. Finally these optical fibers were deployed successfully in a sodium environment for data acquisition. As a test of the sodium deployable optical fiber temperature sensors they were installed in a sub-loop of the sodium facility which was constructed to promote the thermal striping effect in sodium. The optical fibers performed exceptionally well, yielding unprecedented 2 dimensional temperature profiles with good temporal resolution. Finally, this thermal striping loop was used to perform cross correlation velocimetry successfully over a wide range of flow rates.« less