Energy research: accelerator builders eager to aid fusion work.

PubMed

Metz, W D

1976-10-15

Useful fusion energy may be generated by means of heavy ion accelerator driven implosions if the contraints dictated by the physics and economics of thermonuclear targets and reactors can be satisfied.

Effect of high Z impurities on the ignition and Lawson conditions for a thermonuclear reactor

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

Meade, D.M.

1973-06-01

The recent advances in plasma heating and confinement using Tokamak devices have produced plasmas which approach thermonuclear conditions. Substantial amounts (0.1 to 1%) of partially stripped high Z impurities have been observed in these discharges. These high Z impurities (Fe,Mo,W) are presumably due to plasma bombardment of the limiter and vacuum chamber wall. Since the plasma energy will be increasing sharply in the next sequence of experiments from approx. =1kJ in ST tokamak to approx. =3MJ in PLT and up to approx. =100MJ in a feasibility experiment, the bombardment of the wall and limiter will become increasingly important. In thismore » paper, the effects of high Z impurities on the ignition and Lawson conditions for a DT reactor are calculated. 7 refs., 2 figs.« less

Nuclear pursuits

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

Not Available

1993-05-01

This table lists quantities of warheads (in stockpile, peak number per year, total number built, number of known test explosions), weapon development milestones (developers of the atomic bomb and hydrogen bomb, date of first operational ICBM, first nuclear-powered naval SSN in service, first MIRVed missile deployed), and testing milestones (first fission test, type of boosted fission weapon, multistage thermonuclear test, number of months from fission bomb to multistage thermonuclear bomb, etc.), and nuclear infrastructure (assembly plants, plutonium production reactors, uranium enrichment plants, etc.). Countries included in the tally are the United States, Soviet Union, Britain, France, and China.

Monte Carlo simulation of neutral-beam injection for mirror fusion reactors

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

Miller, Ronald Lee

1979-01-01

Computer simulation techniques using the Monte Carlo method have been developed for application to the modeling of neutral-beam intection into mirror-confined plasmas of interest to controlled thermonuclear research. The energetic (10 to 300 keV) neutral-beam particles interact with the target plasma (T i ~ 10 to 100 keV) through electron-atom and ion-atom collisional ionization as well as ion-atom charge-transfer (charge-exchange) collisions to give a fractional trapping of the neutral beam and a loss of charge-transfer-produced neutrals which escape to bombard the reactor first wall. Appropriate interaction cross sections for these processes are calculated for the assumed anisotropic, non-Maxwellian plasma ionmore » phase-space distributions.« less

REACTOR

DOEpatents

Spitzer, L. Jr.

1962-01-01

The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

Current drive at plasma densities required for thermonuclear reactors.

PubMed

Cesario, R; Amicucci, L; Cardinali, A; Castaldo, C; Marinucci, M; Panaccione, L; Santini, F; Tudisco, O; Apicella, M L; Calabrò, G; Cianfarani, C; Frigione, D; Galli, A; Mazzitelli, G; Mazzotta, C; Pericoli, V; Schettini, G; Tuccillo, A A

2010-08-10

Progress in thermonuclear fusion energy research based on deuterium plasmas magnetically confined in toroidal tokamak devices requires the development of efficient current drive methods. Previous experiments have shown that plasma current can be driven effectively by externally launched radio frequency power coupled to lower hybrid plasma waves. However, at the high plasma densities required for fusion power plants, the coupled radio frequency power does not penetrate into the plasma core, possibly because of strong wave interactions with the plasma edge. Here we show experiments performed on FTU (Frascati Tokamak Upgrade) based on theoretical predictions that nonlinear interactions diminish when the peripheral plasma electron temperature is high, allowing significant wave penetration at high density. The results show that the coupled radio frequency power can penetrate into high-density plasmas due to weaker plasma edge effects, thus extending the effective range of lower hybrid current drive towards the domain relevant for fusion reactors.

Effect of particle pinch on the fusion performance and profile features of an international thermonuclear experimental reactor-like fusion reactor

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

Wang, Shijia, E-mail: wangsg@mail.ustc.edu.cn; Wang, Shaojie

2015-04-15

The evolution of the plasma temperature and density in an international thermonuclear experimental reactor (ITER)-like fusion device has been studied by numerically solving the energy transport equation coupled with the particle transport equation. The effect of particle pinch, which depends on the magnetic curvature and the safety factor, has been taken into account. The plasma is primarily heated by the alpha particles which are produced by the deuterium-tritium fusion reactions. A semi-empirical method, which adopts the ITERH-98P(y,2) scaling law, has been used to evaluate the transport coefficients. The fusion performances (the fusion energy gain factor, Q) similar to the ITERmore » inductive scenario and non-inductive scenario (with reversed magnetic shear) are obtained. It is shown that the particle pinch has significant effects on the fusion performance and profiles of a fusion reactor. When the volume-averaged density is fixed, particle pinch can lower the pedestal density by ∼30%, with the Q value and the central pressure almost unchanged. When the particle source or the pedestal density is fixed, the particle pinch can significantly enhance the Q value by  60%, with the central pressure also significantly raised.« less

CuPb rheocast alloy as joining material for CFC composites

NASA Astrophysics Data System (ADS)

Salvo, M.; Lemoine, P.; Ferraris, M.; Appendino Montorsi, M.; Matera, R.

1995-10-01

High heat flux components for future use in thermonuclear fusion reactors are designed as layered structures. The assembling of the different parts (armour, heat sink and external structure) requires a joint which could withstand large heat loads and thermal stresses. In this paper we examined a 50 wt% PbCu rheocast alloy (RCA) as joining material for the armour/heat sink joint. The alloy was prepared in vacuum in a rotational furnace and was characterized by SEM-EDS analysis and heating microscopy. The obtained microstructure was globular as foreseen and it remained after prolonged heating at 650°C. The alloy showed very good ductility: sheets of about 200 μm were rolled starting from about 1 × 1 × 1 cm 3 cubes. The alloy was successful in joining both the armour and the heat sink materials, respectively, carbon fibre reinforced composites and copper. Initial mechanical testing shows that the technique is viable for the foreseen applications in the field of thermonuclear fusion reactors.

Control of Warm Compression Stations Using Model Predictive Control: Simulation and Experimental Results

NASA Astrophysics Data System (ADS)

Bonne, F.; Alamir, M.; Bonnay, P.

2017-02-01

This paper deals with multivariable constrained model predictive control for Warm Compression Stations (WCS). WCSs are subject to numerous constraints (limits on pressures, actuators) that need to be satisfied using appropriate algorithms. The strategy is to replace all the PID loops controlling the WCS with an optimally designed model-based multivariable loop. This new strategy leads to high stability and fast disturbance rejection such as those induced by a turbine or a compressor stop, a key-aspect in the case of large scale cryogenic refrigeration. The proposed control scheme can be used to achieve precise control of pressures in normal operation or to avoid reaching stopping criteria (such as excessive pressures) under high disturbances (such as a pulsed heat load expected to take place in future fusion reactors, expected in the cryogenic cooling systems of the International Thermonuclear Experimental Reactor ITER or the Japan Torus-60 Super Advanced fusion experiment JT-60SA). The paper details the simulator used to validate this new control scheme and the associated simulation results on the SBTs WCS. This work is partially supported through the French National Research Agency (ANR), task agreement ANR-13-SEED-0005.

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

None

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

On the implementation of a chain nuclear reaction of thermonuclear fusion on the basis of the p+11B process

NASA Astrophysics Data System (ADS)

Belyaev, V. S.; Krainov, V. P.; Zagreev, B. V.; Matafonov, A. P.

2015-07-01

Various theoretical and experimental schemes for implementing a thermonuclear reactor on the basis of the p+11B reaction are considered. They include beam collisions, fusion in degenerate plasmas, ignition upon plasma acceleration by ponderomotive forces, and the irradiation of a solid-state target from 11B with a proton beam under conditions of a Coulomb explosion of hydrogen microdrops. The possibility of employing ultra-short high-intensity laser pulses to initiate the p+11B reaction under conditions far from thermodynamic equilibrium is discussed. This and some other weakly radioactive thermonuclear reactions are promising owing to their ecological cleanness—there are virtually no neutrons among fusion products. Nuclear reactions that follow the p+11B reaction may generate high-energy protons, sustaining a chain reaction, and this is an advantage of the p+11B option. The approach used also makes it possible to study nuclear reactions under conditions close to those in the early Universe or in the interior of stars.

Shock ignition: a new approach to high gain inertial confinement fusion on the national ignition facility.

PubMed

Perkins, L J; Betti, R; LaFortune, K N; Williams, W H

2009-07-24

Shock ignition, an alternative concept for igniting thermonuclear fuel, is explored as a new approach to high gain, inertial confinement fusion targets for the National Ignition Facility (NIF). Results indicate thermonuclear yields of approximately 120-250 MJ may be possible with laser drive energies of 1-1.6 MJ, while gains of approximately 50 may still be achievable at only approximately 0.2 MJ drive energy. The scaling of NIF energy gain with laser energy is found to be G approximately 126E (MJ);{0.510}. This offers the potential for high-gain targets that may lead to smaller, more economic fusion power reactors and a cheaper fusion energy development path.

Current situation: New enthusiasm. [Nuclear fusion

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

Not Available

For decades the costly field of controlled nuclear fusion has been rocked by ups and downs, promise and problems. In spite of the many setbacks, scientists and DOE officials are determined to push ahead. [open quotes]We are very confident that by some time after the first decade of the next century, we will have a clear demonstration [of the technology] to give us unlimited energy....We are very excited about it,[close quotes] Energy Secretary Watkins said last spring in proposing a $360 million fusion energy budget for fiscal 1993. This article cites recent hey developments in terms of technical accomplishments, fundingmore » decisions, policy decisions, and efforts to collaborate internationally on controlled nuclear fusion. The International Thermonuclear Experimental Reactor is discussed also.« less

Eye-in-Hand Manipulation for Remote Handling: Experimental Setup

NASA Astrophysics Data System (ADS)

Niu, Longchuan; Suominen, Olli; Aref, Mohammad M.; Mattila, Jouni; Ruiz, Emilio; Esque, Salvador

2018-03-01

A prototype for eye-in-hand manipulation in the context of remote handling in the International Thermonuclear Experimental Reactor (ITER)1 is presented in this paper. The setup consists of an industrial robot manipulator with a modified open control architecture and equipped with a pair of stereoscopic cameras, a force/torque sensor, and pneumatic tools. It is controlled through a haptic device in a mock-up environment. The industrial robot controller has been replaced by a single industrial PC running Xenomai that has a real-time connection to both the robot controller and another Linux PC running as the controller for the haptic device. The new remote handling control environment enables further development of advanced control schemes for autonomous and semi-autonomous manipulation tasks. This setup benefits from a stereovision system for accurate tracking of the target objects with irregular shapes. The overall environmental setup successfully demonstrates the required robustness and precision that remote handling tasks need.

Macromolecular Networks Containing Fluorinated Cyclic Moieties

DTIC Science & Technology

2015-12-12

Approved for public release.  Distribution is unlimited.   Cyanate Esters Around the Solar System 4 Images:  courtesy  NASA  (public release) • The...science decks on the Mars Phoenix lander are made from M55J/cyanate ester composites • The solar panel supports on the MESSENGER space probe use cyanate...thermonuclear fusion reactor Fusion reactor, photo  courtesy of Gerritse ((Wikimedia Commons) • Unique cyanate ester composites have been designed by NASA

The 14 MeV Neutron Irradiation Facility in MARIA Reactor

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

Prokopowicz, R.; Pytel, K.; Dorosz, M.

2015-07-01

The MARIA reactor with thermal neutron flux density up to 3x10{sup 14} cm{sup -2} s{sup -1} and a number of vertical channels is well suited to material testing by thermal neutron treatment. Beside of that some fast neutron irradiation facilities are operated in MARIA reactor as well. One of them is thermal to 14 MeV neutron converter launched in 2014. It is especially devoted to fusion devices material testing irradiation. The ITER and DEMO research thermonuclear facilities are to be run using the deuterium - tritium fusion reaction. Fast neutrons (of energy approximately 14 MeV) resulting from the reaction aremore » essential to carry away the released thermonuclear energy and to breed tritium. However, constructional materials of which thermonuclear reactors are to be built must be specially selected to survive intense fluxes of fast neutrons. Strong sources of 14 MeV neutrons are needed if research on resistance of candidate materials to such fluxes is to be carried out effectively. Nuclear reactor-based converter capable to convert thermal neutrons into 14 MeV fast neutrons may be used to that purpose. The converter based on two stage nuclear reaction on lithium-6 and deuterium compounds leading to 14 MeV neutron production. The reaction chain is begun by thermal neutron capture by lithium-6 nucleus resulted in triton release. The neutron and triton transport calculations have been therefore carried-out to estimate the thermal to 14 MeV neutron conversion efficiency and optimize converter construction. The usable irradiation space of ca. 60 cm{sup 3} has been obtained. The released energy have been calculated. Heat transport has been asses to ensure proper device cooling. A set of thermocouples has been installed in converter to monitor its temperature distribution on-line. Influence of converter on reactor operation has been studied. Safety analyses of steady states and transients have been done. Performed calculations and analyses allow designing the converter and formulate its operation limits and conditions. During first tested operation of the converter the 14 MeV neutron flux density was estimated to 10{sup 9} cm{sup -2} s{sup -1}, whereas fast fission neutrons inside converter achieved 10{sup 12} cm{sup -2} s{sup -1}, and thermal neutrons were reduced down to 109 cm-2 s-1. Taking into account the feasibility of almost incessant converter operation for a number of months, its arisen as one of the most powerful (in terms of fluence), currently available 14 MeV neutron source. Such a converter currently under operation in the MARIA reactor core will be presented. (authors)« less

EFFECTS OF LASER RADIATION ON MATTER. LASER PLASMA: Physics of the plasma corona in the problem of laser controlled thermonuclear fusion

NASA Astrophysics Data System (ADS)

Andreev, N. E.; Gorbunov, Leonid M.; Tikhonchuk, Vladimir T.

1994-09-01

A brief analysis is made of the most important nonlinear processes which result from the interaction of laser radiation with thermonuclear targets. lt is shown that problems in the physics of the plasma corona should be an essential part of any programme of research on laser controlled thermonuclear fusion. A list is given of the problems that have to be solved first before going to the next level of laser energies.

On the implementation of a chain nuclear reaction of thermonuclear fusion on the basis of the p+{sup 11}B process

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

Belyaev, V. S.; Krainov, V. P., E-mail: vpkrainov@mail.ru; Zagreev, B. V.

2015-07-15

Various theoretical and experimental schemes for implementing a thermonuclear reactor on the basis of the p+{sup 11}B reaction are considered. They include beam collisions, fusion in degenerate plasmas, ignition upon plasma acceleration by ponderomotive forces, and the irradiation of a solid-state target from {sup 11}B with a proton beam under conditions of a Coulomb explosion of hydrogen microdrops. The possibility of employing ultra-short high-intensity laser pulses to initiate the p+{sup 11}B reaction under conditions far from thermodynamic equilibrium is discussed. This and some other weakly radioactive thermonuclear reactions are promising owing to their ecological cleanness—there are virtually no neutrons amongmore » fusion products. Nuclear reactions that follow the p+{sup 11}B reaction may generate high-energy protons, sustaining a chain reaction, and this is an advantage of the p+{sup 11}B option. The approach used also makes it possible to study nuclear reactions under conditions close to those in the early Universe or in the interior of stars.« less

Neoclassical Theory and Its Applications

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

Shaing, Ker-Chung

2015-11-20

The grant entitled Neoclassical Theory and Its Applications started on January 15 2001 and ended on April 14 2015. The main goal of the project is to develop neoclassical theory to understand tokamak physics, and employ it to model current experimental observations and future thermonuclear fusion reactors. The PI had published more than 50 papers in refereed journals during the funding period.

Application of microdosimetry on biological physics for ionizing radiation

NASA Astrophysics Data System (ADS)

Chen, Dandan; Sun, Liang

2018-02-01

Not Available Project supported by the National Natural Science Foundation of China (Grant Nos. 11304212 and 11575124), the Natural Science Foundation of Jiangsu Province, China (Grant Nos. BK20130279), the Priority Academic Program Development of Jiangsu Higher Education Institutions (PAPD), and the International Thermonuclear Experimental Reactor (ITER) Special Program of China (Grant No. 2014GB112006).

Telescope-based cavity for negative ion beam neutralization in future fusion reactors.

PubMed

Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

2018-03-01

In future fusion reactors, heating system efficiency is of the utmost importance. Photo-neutralization substantially increases the neutral beam injector (NBI) efficiency with respect to the foreseen system in the International Thermonuclear Experimental Reactor (ITER) based on a gaseous target. In this paper, we propose a telescope-based configuration to be used in the NBI photo-neutralizer cavity of the demonstration power plant (DEMO) project. This configuration greatly reduces the total length of the cavity, which likely solves overcrowding issues in a fusion reactor environment. Brought to a tabletop experiment, this cavity configuration is tested: a 4 mm beam width is obtained within a ≃1.5  m length cavity. The equivalent cavity g factor is measured to be 0.038(3), thus confirming the cavity stability.

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.

Neutral atom analyzers for diagnosing hot plasmas: A review of research at the ioffe physicotechnical institute

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

Kislyakov, A. I.; Petrov, M. P.

2009-07-15

Research on neutral particle diagnostics of thermonuclear plasmas that has been carried out in recent years at the Ioffe Physicotechnical Institute of the Russian Academy of Sciences (St. Petersburg, Russia) is reviewed. Work on the creation and improvement of neutral atom analyzers was done in two directions: for potential applications (in particular, on the International Thermonuclear Experimental Reactor, which is now under construction at Cadarache in France) and for investigation of the ion plasma component in various devices (in particular, in the largest tokamaks, such as JET, TFTR, and JT-60). Neutral atom analyzers are the main tool for studying themore » behavior of hydrogen ions and isotopes in magnetic confinement systems. They make it possible to determine energy spectra, to perform the isotope analysis of atom fluxes from the plasma, to measure the absolute intensity of the fluxes, and to record how these parameters vary with time. A comparative description of the analyzers developed in recent years at the Ioffe Institute is given. These are ACORD-12/24 analyzers for recording 0.2-100-keV hydrogen and deuterium atoms with a tunable range of simultaneously measured energies, CNPA compact analyzers for a fixed energy gain in the ranges 80-1000 eV and 0.8-100 keV, an ISEP analyzer for simultaneously recording the atoms of all the three hydrogen isotopes (H, D, and T) in the energy range 5-700 keV, and GEMMA analyzers for recording atom fluxes of hydrogen and helium isotopes in the range 0.1-4 MeV. The scintillating detectors of the ISEP and GEMMA analyzers have a lowered sensitivity to neutrons and thus can operate without additional shielding in neutron fields of up to 10{sup 9} n/(cm{sup 2} s). These two types of analyzers, intended to operate under deuterium-tritium plasma conditions, are prototypes of atom analyzers created at the Ioffe Institute for use in the International Thermonuclear Experimental Reactor. With these analyzers, a number of new results have been obtained in recent years in various devices. Some results are presented from investigation of ions in the Globus-M spherical tokamak, the W7-AS stellarator, and the JET tokamak by means of the analyzers developed at the Ioffe Institute. Challenges and opportunities for applying these diagnostics in the International Thermonuclear Experimental Reactor project are discussed.« less

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.

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

DOEpatents

Lasche, G.P.

1983-09-29

The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

NASA Astrophysics Data System (ADS)

Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

2017-12-01

The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

First operation with the JET International Thermonuclear Experimental Reactor-like walla)

NASA Astrophysics Data System (ADS)

Neu, R.; Arnoux, G.; Beurskens, M.; Bobkov, V.; Brezinsek, S.; Bucalossi, J.; Calabro, G.; Challis, C.; Coenen, J. W.; de la Luna, E.; de Vries, P. C.; Dux, R.; Frassinetti, L.; Giroud, C.; Groth, M.; Hobirk, J.; Joffrin, E.; Lang, P.; Lehnen, M.; Lerche, E.; Loarer, T.; Lomas, P.; Maddison, G.; Maggi, C.; Matthews, G.; Marsen, S.; Mayoral, M.-L.; Meigs, A.; Mertens, Ph.; Nunes, I.; Philipps, V.; Pütterich, T.; Rimini, F.; Sertoli, M.; Sieglin, B.; Sips, A. C. C.; van Eester, D.; van Rooij, G.; JET-EFDA Contributors

2013-05-01

To consolidate International Thermonuclear Experimental Reactor (ITER) design choices and prepare for its operation, Joint European Torus (JET) has implemented ITER's plasma facing materials, namely, Be for the main wall and W in the divertor. In addition, protection systems, diagnostics, and the vertical stability control were upgraded and the heating capability of the neutral beams was increased to over 30 MW. First results confirm the expected benefits and the limitations of all metal plasma facing components (PFCs) but also yield understanding of operational issues directly relating to ITER. H-retention is lower by at least a factor of 10 in all operational scenarios compared to that with C PFCs. The lower C content (≈ factor 10) has led to much lower radiation during the plasma burn-through phase eliminating breakdown failures. Similarly, the intrinsic radiation observed during disruptions is very low, leading to high power loads and to a slow current quench. Massive gas injection using a D2/Ar mixture restores levels of radiation and vessel forces similar to those of mitigated disruptions with the C wall. Dedicated L-H transition experiments indicate a 30% power threshold reduction, a distinct minimum density, and a pronounced shape dependence. The L-mode density limit was found to be up to 30% higher than for C allowing stable detached divertor operation over a larger density range. Stable H-modes as well as the hybrid scenario could be re-established only when using gas puff levels of a few 1021 es-1. On average, the confinement is lower with the new PFCs, but nevertheless, H factors up to 1 (H-Mode) and 1.3 (at βN≈3, hybrids) have been achieved with W concentrations well below the maximum acceptable level.

A Simulink Library of cryogenic components to automatically generate control schemes for large Cryorefrigerators

NASA Astrophysics Data System (ADS)

Bonne, François; Alamir, Mazen; Hoa, Christine; Bonnay, Patrick; Bon-Mardion, Michel; Monteiro, Lionel

2015-12-01

In this article, we present a new Simulink library of cryogenics components (such as valve, phase separator, mixer, heat exchanger...) to assemble to generate model-based control schemes. Every component is described by its algebraic or differential equation and can be assembled with others to build the dynamical model of a complete refrigerator or the model of a subpart of it. The obtained model can be used to automatically design advanced model based control scheme. It also can be used to design a model based PI controller. Advanced control schemes aim to replace classical user experience designed approaches usually based on many independent PI controllers. This is particularly useful in the case where cryoplants are submitted to large pulsed thermal loads, expected to take place in future fusion reactors such as those expected in the cryogenic cooling systems of the International Thermonuclear Experimental Reactor (ITER) or the Japan Torus-60 Super Advanced Fusion Experiment (JT- 60SA). The paper gives the example of the generation of the dynamical model of the 400W@1.8K refrigerator and shows how to build a Constrained Model Predictive Control for it. Based on the scheme, experimental results will be given. This work is being supported by the French national research agency (ANR) through the ANR-13-SEED-0005 CRYOGREEN program.

History of 232-F, tritium extraction processing

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

Blackburn, G.W.

1994-08-01

In 1950 the Atomic Energy Commission authorized the Savannah River Project principally for the production of tritium and plutonium-239 for use in thermonuclear weapons. 232-F was built as an interim facility in 1953--1954, at a cost of $3.9M. Tritium extraction operations began in October, 1955, after the reactor and separations startups. In July, 1957 a larger tritium facility began operation in 232-H. In 1958 the capacity of 232-H was doubled. Also, in 1957 a new task was assigned to Savannah River, the loading of tritium into reservoirs that would be actual components of thermonuclear weapons. This report describes the historymore » of 232-F, the process for tritium extraction, and the lessons learned over the years that were eventually incorporated into the new Replacement Tritium Facility.« less

Three-dimensional analysis of tokamaks and stellarators

PubMed Central

Garabedian, Paul R.

2008-01-01

The NSTAB equilibrium and stability code and the TRAN Monte Carlo transport code furnish a simple but effective numerical simulation of essential features of present tokamak and stellarator experiments. When the mesh size is comparable to the island width, an accurate radial difference scheme in conservation form captures magnetic islands successfully despite a nested surface hypothesis imposed by the mathematics. Three-dimensional asymmetries in bifurcated numerical solutions of the axially symmetric tokamak problem are relevant to the observation of unstable neoclassical tearing modes and edge localized modes in experiments. Islands in compact stellarators with quasiaxial symmetry are easier to control, so these configurations will become good candidates for magnetic fusion if difficulties with safety and stability are encountered in the International Thermonuclear Experimental Reactor (ITER) project. PMID:18768807

Beryllium processing technology review for applications in plasma-facing components

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

Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

1993-07-01

Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itselfmore » and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included.« less

Thermonuclear inverse magnetic pumping power cycle for stellarator reactor

DOEpatents

Ho, Darwin D.; Kulsrud, Russell M.

1991-01-01

The plasma column in a stellarator is compressed and expanded alternatively in minor radius. First a plasma in thermal balance is compressed adiabatically. The volume of the compressed plasma is maintained until the plasma reaches a new thermal equilibrium. The plasma is then expanded to its original volume. As a result of the way a stellarator works, the plasma pressure during compression is less than the corresponding pressure during expansion. Therefore, negative work is done on the plasma over a complete cycle. This work manifests itself as a back-voltage in the toroidal field coils. Direct electrical energy is obtained from this voltage. Alternatively, after the compression step, the plasma can be expanded at constant pressure. The cycle can be made self-sustaining by operating a system of two stellarator reactors in tandem. Part of the energy derived from the expansion phase of a first stellarator reactor is used to compress the plasma in a second stellarator reactor.

Hybrid fusion-fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

NASA Astrophysics Data System (ADS)

Shmelev, A. N.; Kulikov, G. G.; Kurnaev, V. A.; Salahutdinov, G. H.; Kulikov, E. G.; Apse, V. A.

2015-12-01

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the 231Pa-232U-233U-Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of 232U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.

APPARATUS FOR HEATING A PLASMA

DOEpatents

Stix, T.H.

1962-01-01

The system contemplates the use of ion cyclotron motions for transferring energy to a plasma immersed in a confining magnetic field such as is found in thermonuclear reactors of the stellarator class. Oppositely directed windings are provided for producing ion-accelerating fields having a time and spatial periodicity and these have the advantage of producing ion cyclotron motions without the development of space charges which preclude the efficient energy transfer to the plasma. (AEC)

Beryllium fabrication/cost assessment for ITER (International Thermonuclear Experimental Reactor)

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

Beeston, J.M.; Longhurst, G.R.; Parsonage, T.

1990-06-01

A fabrication and cost estimate of three possible beryllium shapes for the International Thermonuclear Experimental Reactor (ITER) blanket is presented. The fabrication method by hot pressing (HP), cold isostatic pressing plus sintering (CIP+S), cold isostatic pressing plus sintering plus hot isostatic pressing (CIP+S+HIP), and sphere production by atomization or rotary electrode will be discussed. Conventional hot pressing blocks of beryllium with subsequent machining to finished shapes can be more expensive than production of a net shape by cold isostatic pressing and sintering. The three beryllium shapes to be considered here and proposed for ITER are: (1) cubic blocks (3 tomore » 17 cm on an edge), (2) tubular cylinders (33 to 50 mm i.d. by 62 mm o.d. by 8 m long), and (3) spheres (1--5 mm dia.). A rough cost estimate of the basic shape is presented which would need to be refined if the surface finish and tolerances required are better than the sintering process produces. The final cost of the beryllium in the blanket will depend largely on the machining and recycling of beryllium required to produce the finished product. The powder preparation will be discussed before shape fabrication. 10 refs., 6 figs.« less

Method of achieving the controlled release of thermonuclear energy

DOEpatents

Brueckner, Keith A.

1986-01-01

A method of achieving the controlled release of thermonuclear energy by illuminating a minute, solid density, hollow shell of a mixture of material such as deuterium and tritium with a high intensity, uniformly converging laser wave to effect an extremely rapid build-up of energy in inwardly traveling shock waves to implode the shell creating thermonuclear conditions causing a reaction of deuterons and tritons and a resultant high energy thermonuclear burn. Utilizing the resulting energy as a thermal source and to breed tritium or plutonium. The invention also contemplates a laser source wherein the flux level is increased with time to reduce the initial shock heating of fuel and provide maximum compression after implosion; and, in addition, computations and an equation are provided to enable the selection of a design having a high degree of stability and a dependable fusion performance by establishing a proper relationship between the laser energy input and the size and character of the selected material for the fusion capsule.

The Joint European Torus (JET)

NASA Astrophysics Data System (ADS)

Rebut, Paul-Henri

2017-02-01

This paper addresses the history of JET, the Tokamak that reached the highest performances and the experiment that so far came closest to the eventual goal of a fusion reactor. The reader must be warned, however, that this document is not a comprehensive study of controlled thermonuclear fusion or even of JET. The next step on this road, the ITER project, is an experimental reactor. Actually, several prototypes will be required before a commercial reactor can be built. The fusion history is far from been finalised. JET is still in operation some 32 years after the first plasma and still has to provide answers to many questions before ITER takes the lead on research. Some physical interpretations of the observed phenomena, although coherent, are still under discussion. This paper also recalls some basic physics concepts necessary to the understanding of confinement: a knowledgeable reader can ignore these background sections. This fascinating story, comprising successes and failures, is imbedded in the complexities of twentieth and the early twenty-first centuries at a time when world globalization is evolving and the future seems loaded with questions. The views here expressed on plasma confinement are solely those of the author. This is especially the case for magnetic turbulence, for which other scientists may have different views.

Current status of nuclear engineering education

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

Palladino, N.J.

1975-09-01

The 65 colleges and universities offering undergraduate degrees in nuclear engineering and the 15 schools offering strong nuclear engineering options are, in general, doing a good job to meet the current spectrum of job opportunities. But, nuclear engineering programs are not producing enough graduates to meet growing demands. They currently receive little aid and support from their customers --industry and government--in the form of scholarships, grants, faculty research support, student thesis and project support, or student summer jobs. There is not enough interaction between industry and universities. Most nuclear engineering programs are geared too closely to the technology of themore » present family of reactors and too little to the future breeder reactors and controlled thermonuclear reactors. In addition, nuclear engineering programs attract too few women and members of minority ethnic groups. Further study of the reasons for this fact is needed so that effective corrective action can be taken. Faculty in nuclear engineering programs should assume greater initiative to provide attractive and objective nuclear energy electives for technical and nontechnical students in other disciplines to improve their technical understanding of the safety and environmental issues involved. More aggressive and persistent efforts must be made by nuclear engineering schools to obtain industry support and involvement in their programs. (auth)« less

Physical control oriented model of large scale refrigerators to synthesize advanced control schemes. Design, validation, and first control results

NASA Astrophysics Data System (ADS)

Bonne, François; Alamir, Mazen; Bonnay, Patrick

2014-01-01

In this paper, a physical method to obtain control-oriented dynamical models of large scale cryogenic refrigerators is proposed, in order to synthesize model-based advanced control schemes. These schemes aim to replace classical user experience designed approaches usually based on many independent PI controllers. This is particularly useful in the case where cryoplants are submitted to large pulsed thermal loads, expected to take place in the cryogenic cooling systems of future fusion reactors such as the International Thermonuclear Experimental Reactor (ITER) or the Japan Torus-60 Super Advanced Fusion Experiment (JT-60SA). Advanced control schemes lead to a better perturbation immunity and rejection, to offer a safer utilization of cryoplants. The paper gives details on how basic components used in the field of large scale helium refrigeration (especially those present on the 400W @1.8K helium test facility at CEA-Grenoble) are modeled and assembled to obtain the complete dynamic description of controllable subsystems of the refrigerator (controllable subsystems are namely the Joule-Thompson Cycle, the Brayton Cycle, the Liquid Nitrogen Precooling Unit and the Warm Compression Station). The complete 400W @1.8K (in the 400W @4.4K configuration) helium test facility model is then validated against experimental data and the optimal control of both the Joule-Thompson valve and the turbine valve is proposed, to stabilize the plant under highly variable thermals loads. This work is partially supported through the European Fusion Development Agreement (EFDA) Goal Oriented Training Program, task agreement WP10-GOT-GIRO.

Physical control oriented model of large scale refrigerators to synthesize advanced control schemes. Design, validation, and first control results

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

Bonne, François; Bonnay, Patrick; Alamir, Mazen

2014-01-29

In this paper, a physical method to obtain control-oriented dynamical models of large scale cryogenic refrigerators is proposed, in order to synthesize model-based advanced control schemes. These schemes aim to replace classical user experience designed approaches usually based on many independent PI controllers. This is particularly useful in the case where cryoplants are submitted to large pulsed thermal loads, expected to take place in the cryogenic cooling systems of future fusion reactors such as the International Thermonuclear Experimental Reactor (ITER) or the Japan Torus-60 Super Advanced Fusion Experiment (JT-60SA). Advanced control schemes lead to a better perturbation immunity and rejection,more » to offer a safer utilization of cryoplants. The paper gives details on how basic components used in the field of large scale helium refrigeration (especially those present on the 400W @1.8K helium test facility at CEA-Grenoble) are modeled and assembled to obtain the complete dynamic description of controllable subsystems of the refrigerator (controllable subsystems are namely the Joule-Thompson Cycle, the Brayton Cycle, the Liquid Nitrogen Precooling Unit and the Warm Compression Station). The complete 400W @1.8K (in the 400W @4.4K configuration) helium test facility model is then validated against experimental data and the optimal control of both the Joule-Thompson valve and the turbine valve is proposed, to stabilize the plant under highly variable thermals loads. This work is partially supported through the European Fusion Development Agreement (EFDA) Goal Oriented Training Program, task agreement WP10-GOT-GIRO.« less

Reaching to a featured formula to deduce the energy of the heaviest particles producing from the controlled thermonuclear fusion reactions

NASA Astrophysics Data System (ADS)

Majeed, Raad H.; Oudah, Osamah N.

2018-05-01

Thermonuclear fusion reaction plays an important role in developing and construction any power plant system. Studying the physical behavior for the possible mechanism governed energies released by the fusion products to precise understanding the related kinematics. In this work a theoretical formula controlled the general applied thermonuclear fusion reactions is achieved to calculating the fusion products energy depending upon the reactants physical properties and therefore, one can calculate other parameters governed a given reaction. By using this formula, the energy spectrum of 4He produced from T-3He fusion reaction has been sketched with respect to reaction angle and incident energy ranged from (0.08-0.6) MeV.

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

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

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of themore » important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.« less

Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

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

Shmelev, A. N., E-mail: shmelan@mail.ru; Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Kurnaev, V. A., E-mail: kurnaev@yandex.ru

2015-12-15

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the {sup 231}Pa–{sup 232}U–{sup 233}U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be bettermore » protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of {sup 232}U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.« less

Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

NASA Astrophysics Data System (ADS)

Poddubnyi, I. I.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V.; Leshukov, A. Yu.; Aleskovskiy, K. V.; Obukhov, D. M.

2016-12-01

The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational convection (TGC) on averaged and fluctuating quantities were found. The development of TGC in the presence of a magnetic field leads to the appearance of low-frequency fluctuations whose anomalously high intensity exceeds severalfold the level of turbulence fluctuations. This effect manifest itself over a broad region of regime parameters. It was confirmed that low-energy fluctuations penetrate readily through the wall; therefore, it is necessary to study this effect further—in particular, from the point of view of the fatigue strength of the walls of liquid-metal channels.

Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

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

Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru; Pyatnitskaya, N. Yu.; Razuvanov, N. G.

2016-12-15

The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in themore » majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational convection (TGC) on averaged and fluctuating quantities were found. The development of TGC in the presence of a magnetic field leads to the appearance of low-frequency fluctuations whose anomalously high intensity exceeds severalfold the level of turbulence fluctuations. This effect manifest itself over a broad region of regime parameters. It was confirmed that low-energy fluctuations penetrate readily through the wall; therefore, it is necessary to study this effect further—in particular, from the point of view of the fatigue strength of the walls of liquid-metal channels.« less

Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

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

Hoffman, M.A.; Werner, R.W.; Carlson, G.A.

1976-04-01

Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements.more » Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated.« less

Fusion policy advisory committee named

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

Not Available

Department of Energy Secretary James Watkins has announced the formation of new Fusion Policy Advisory Committee which will recommend a policy for conducting DOE's fusion energy research program. Issues that will be considered by the committee include the balance of research activities within the programs, the timing of experiments to test the burning of plasma fuel, the International Thermonuclear Experimental Reactor, and the development of laser technologies, DOE said. Watkins said that he would be entirely open to the committee's advice.

Direct conversion of nuclear radiation energy

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

Miley, George H.

1970-01-01

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

Process for recovering tritium from molten lithium metal

DOEpatents

Maroni, Victor A.

1976-01-01

Lithium tritide (LiT) is extracted from molten lithium metal that has been exposed to neutron irradiation for breeding tritium within a thermonuclear or fission reactor. The extraction is performed by intimately contacting the molten lithium metal with a molten lithium salt, for instance, lithium chloride - potassium chloride eutectic to distribute LiT between the salt and metal phases. The extracted tritium is recovered in gaseous form from the molten salt phase by a subsequent electrolytic or oxidation step.

Divertor impurity monitor for the International Thermonuclear Experimental Reactor

NASA Astrophysics Data System (ADS)

Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

1999-01-01

The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λ<450 nm) to prevent the transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

A Case Study of Modern PLC and LabVIEW Controls: Power Supply Controls for the ORNL ITER ECH Test Stand

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

Barker, Alan M; Killough, Stephen M; Bigelow, Tim S

2011-01-01

Power Supply Controls are being developed at Oak Ridge National Laboratory (ORNL) to test transmission line components of the Electron Cyclotron Heating (ECH) system, with a focus on gyrotrons and waveguides, in support of the International Thermonuclear Experimental Reactor (ITER). The control is performed by several Programmable Logic Controllers (PLC s) located near the different equipment. A technique of Supervisory Control and Data Acquisition (SCADA) is presented to monitor, control, and log actions of the PLC s on a PC through use of Allen Bradley s Remote I/O communication interface coupled with an Open Process Control/Object Linking and Embedding [OLE]more » for Process Control (OPC) Server/Client architecture. The OPC data is then linked to a National Instruments (NI) LabVIEW system for monitoring and control. Details of the architecture and insight into applicability to other systems are presented in the rest of this paper. Future integration with an EPICS (Experimental Physics Industrial Control System) based mini-CODAC (Control, Data Access and Communication) SCADA system is under consideration, and integration considerations will be briefly introduced.« 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

Advanced low-activation materials. Fibre-reinforced ceramic composites

NASA Astrophysics Data System (ADS)

Fenici, P.; Scholz, H. W.

1994-09-01

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

Model based multivariable controller for large scale compression stations. Design and experimental validation on the LHC 18KW cryorefrigerator

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

Bonne, François; Bonnay, Patrick; Alamir, Mazen

2014-01-29

In this paper, a multivariable model-based non-linear controller for Warm Compression Stations (WCS) is proposed. The strategy is to replace all the PID loops controlling the WCS with an optimally designed model-based multivariable loop. This new strategy leads to high stability and fast disturbance rejection such as those induced by a turbine or a compressor stop, a key-aspect in the case of large scale cryogenic refrigeration. The proposed control scheme can be used to have precise control of every pressure in normal operation or to stabilize and control the cryoplant under high variation of thermal loads (such as a pulsedmore » heat load expected to take place in future fusion reactors such as those expected in the cryogenic cooling systems of the International Thermonuclear Experimental Reactor ITER or the Japan Torus-60 Super Advanced fusion experiment JT-60SA). The paper details how to set the WCS model up to synthesize the Linear Quadratic Optimal feedback gain and how to use it. After preliminary tuning at CEA-Grenoble on the 400W@1.8K helium test facility, the controller has been implemented on a Schneider PLC and fully tested first on the CERN's real-time simulator. Then, it was experimentally validated on a real CERN cryoplant. The efficiency of the solution is experimentally assessed using a reasonable operating scenario of start and stop of compressors and cryogenic turbines. This work is partially supported through the European Fusion Development Agreement (EFDA) Goal Oriented Training Program, task agreement WP10-GOT-GIRO.« less

Model based multivariable controller for large scale compression stations. Design and experimental validation on the LHC 18KW cryorefrigerator

NASA Astrophysics Data System (ADS)

Bonne, François; Alamir, Mazen; Bonnay, Patrick; Bradu, Benjamin

2014-01-01

In this paper, a multivariable model-based non-linear controller for Warm Compression Stations (WCS) is proposed. The strategy is to replace all the PID loops controlling the WCS with an optimally designed model-based multivariable loop. This new strategy leads to high stability and fast disturbance rejection such as those induced by a turbine or a compressor stop, a key-aspect in the case of large scale cryogenic refrigeration. The proposed control scheme can be used to have precise control of every pressure in normal operation or to stabilize and control the cryoplant under high variation of thermal loads (such as a pulsed heat load expected to take place in future fusion reactors such as those expected in the cryogenic cooling systems of the International Thermonuclear Experimental Reactor ITER or the Japan Torus-60 Super Advanced fusion experiment JT-60SA). The paper details how to set the WCS model up to synthesize the Linear Quadratic Optimal feedback gain and how to use it. After preliminary tuning at CEA-Grenoble on the 400W@1.8K helium test facility, the controller has been implemented on a Schneider PLC and fully tested first on the CERN's real-time simulator. Then, it was experimentally validated on a real CERN cryoplant. The efficiency of the solution is experimentally assessed using a reasonable operating scenario of start and stop of compressors and cryogenic turbines. This work is partially supported through the European Fusion Development Agreement (EFDA) Goal Oriented Training Program, task agreement WP10-GOT-GIRO.

Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

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

Lee C. Cadwallader

2010-06-01

This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

Preliminary Failure Modes and Effects Analysis of the US DCLL Test Blanket Module

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

Lee C. Cadwallader

2007-08-01

This report presents the results of a preliminary failure modes and effects analysis (FMEA) of a small tritium-breeding test blanket module design for the International Thermonuclear Experimental Reactor. The FMEA was quantified with “generic” component failure rate data, and the failure events are binned into postulated initiating event families and frequency categories for safety assessment. An appendix to this report contains repair time data to support an occupational radiation exposure assessment for test blanket module maintenance.

Critical threshold behavior for steady-state internal transport barriers in burning plasmas.

PubMed

García, J; Giruzzi, G; Artaud, J F; Basiuk, V; Decker, J; Imbeaux, F; Peysson, Y; Schneider, M

2008-06-27

Burning tokamak plasmas with internal transport barriers are investigated by means of integrated modeling simulations. The barrier sustainment in steady state, differently from the barrier formation process, is found to be characterized by a critical behavior, and the critical number of the phase transition is determined. Beyond a power threshold, alignment of self-generated and noninductively driven currents occurs and steady state becomes possible. This concept is applied to simulate a steady-state scenario within the specifications of the International Thermonuclear Experimental Reactor.

Epidemiology of a thermonuclear bomb-burst over Nashville, Tennessee: a theoretic study.

PubMed

Quinn, R W

1983-07-01

A thermonuclear bomb explosion over any city in the world would have a devastating effect on the population and environment. For those who survive, with or without injuries, life would become primitive with little or no uncontaminated food or water, and with inadequate housing, fuel, and medical care, resulting in a breakdown of family and interpersonal relationships. This theoretic study of the potential outcome of a thermonuclear bomb-burst over Nashville, Tennessee, discusses epidemiologically the wide range of medical and psychologic effects from the direct trauma of blast and fire, widespread epidemics of otherwise controlled disease, long-term chronic illness, genetic damage, and catastrophic environmental havoc.

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Fusion energy from the Moon for the twenty-first century

NASA Technical Reports Server (NTRS)

Kulcinski, G. L.; Cameron, E. N.; Santarius, J. F.; Sviatoslavsky, I. N.; Wittenberg, L. J.; Schmitt, Harrison H.

1992-01-01

It is shown in this paper that the D-He-3 fusion fuel cycle is not only credible from a physics standpoint, but that its breakeven and ignition characteristics could be developed on roughly the same time schedule as the DT cycle. It was also shown that the extremely low fraction of power in neutrons, the lack of significant radioactivity in the reactants, and the potential for very high conversion efficiencies, can result in definite advantages for the D-He-3 cycle with respect to DT fusion and fission reactors in the twenty-first century. More specifically, the D-He-3 cycle can accomplish the following: (1) eliminate the need for deep geologic waste burial facilities and the wastes can qualify for Class A, near-surface land burial; (2) allow 'inherently safe' reactors to be built that, under the worst conceivable accident, cannot cause a civilian fatality or result in a significant (greater than 100 mrem) exposure to a member of the public; (3) reduce the radiation damage levels to a point where no scheduled replacement of reactor structural components is required, i.e., full reactor lifetimes (approximately 30 FPY) can be credibly claimed; (4) increase the reliability and availability of fusion reactors compared to DT systems because of the greatly reduced radioactivity, the low neutron damage, and the elimination of T breeding; and (5) greatly reduce the capital costs of fusion power plants (compared to DT systems) by as much as 50 percent and present the potential for a significant reduction on the COE. The concepts presented in this paper tie together two of the most ambitious high-technology endeavors of the twentieth century: the development of controlled thermonuclear fusion for civilian power applications and the utilization of outer space for the benefit of mankind on Earth.

Fusion energy from the Moon for the twenty-first century

NASA Astrophysics Data System (ADS)

Kulcinski, G. L.; Cameron, E. N.; Santarius, J. F.; Sviatoslavsky, I. N.; Wittenberg, L. J.; Schmitt, Harrison H.

1992-09-01

It is shown in this paper that the D-He-3 fusion fuel cycle is not only credible from a physics standpoint, but that its breakeven and ignition characteristics could be developed on roughly the same time schedule as the DT cycle. It was also shown that the extremely low fraction of power in neutrons, the lack of significant radioactivity in the reactants, and the potential for very high conversion efficiencies, can result in definite advantages for the D-He-3 cycle with respect to DT fusion and fission reactors in the twenty-first century. More specifically, the D-He-3 cycle can accomplish the following: (1) eliminate the need for deep geologic waste burial facilities and the wastes can qualify for Class A, near-surface land burial; (2) allow 'inherently safe' reactors to be built that, under the worst conceivable accident, cannot cause a civilian fatality or result in a significant (greater than 100 mrem) exposure to a member of the public; (3) reduce the radiation damage levels to a point where no scheduled replacement of reactor structural components is required, i.e., full reactor lifetimes (approximately 30 FPY) can be credibly claimed; (4) increase the reliability and availability of fusion reactors compared to DT systems because of the greatly reduced radioactivity, the low neutron damage, and the elimination of T breeding; and (5) greatly reduce the capital costs of fusion power plants (compared to DT systems) by as much as 50 percent and present the potential for a significant reduction on the COE. The concepts presented in this paper tie together two of the most ambitious high-technology endeavors of the twentieth century: the development of controlled thermonuclear fusion for civilian power applications and the utilization of outer space for the benefit of mankind on Earth.

Synchronized fusion development considering physics, materials and heat transfer

NASA Astrophysics Data System (ADS)

Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

2017-12-01

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

Hydrogen and deuterium transport and inventory parameters through W and W-alloys for fusion reactor applications

NASA Astrophysics Data System (ADS)

Benamati, G.; Serra, E.; Wu, C. H.

2000-12-01

The aim of this work is to measure the hydrogen/deuterium transport and inventory parameters in relevant structural and/or armour materials for the International Thermonuclear Experimental Reactor (ITER) divertor such as W and W-alloys. The W-alloys: W, W + 1% La 2O 3 and W + 5% Re have been investigated. The materials were supplied from the Metallwerk Plansee GmbH (Austria). Measurements were conducted using a time-dependent permeation method over the temperature range 673-873 K with hydrogen and deuterium pressures in the range 10-100 kPa (100-1000 mbar). The samples were also characterized using optical microscopy, SEM and energy dispersive spectroscopy (EDS) in order to investigate the composition, microstructure and morphology of the surfaces and cross-sections through the samples.

Laser-driven fusion reactor

DOEpatents

Hedstrom, J.C.

1973-10-01

A laser-driven fusion reactor consisting of concentric spherical vessels in which the thermonuclear energy is derived from a deuterium-tritium (D + T) burn within a pellet'', located at the center of the vessels and initiated by a laser pulse. The resulting alpha -particle energy and a small fraction of the neutron energy are deposited within the pellet; this pellet energy is eventually transformed into sensible heat of lithium in a condenser outside the vessels. The remaining neutron energy is dissipated in a lithium blanket, located within the concentric vessels, where the fuel ingredient, tritium, is also produced. The heat content of the blanket and of the condenser lithium is eventually transferred to a conventional thermodynamic plant where the thermal energy is converted to electrical energy in a steam Rankine cycle. (Official Gazette)

Human machine interface to manually drive rhombic like vehicles in remote handling operations

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

Lopes, Pedro; Vale, Alberto; Ventura, Rodrigo

2015-07-01

In the thermonuclear experimental reactor ITER, a vehicle named CTS is designed to transport a container with activated components inside the buildings. In nominal operations, the CTS is autonomously guided under supervision. However, in some unexpected situations, such as in rescue and recovery operations, the autonomous mode must be overridden and the CTS must be remotely guided by an operator. The CTS is a rhombic-like vehicle, with two drivable and steerable wheels along its longitudinal axis, providing omni-directional capabilities. The rhombic kinematics correspond to four control variables, which are difficult to manage in manual mode operation. This paper proposes amore » Human Machine Interface (HMI) to remotely guide the vehicle in manual mode. The proposed solution is implemented using a HMI with an encoder connected to a micro-controller and an analog 2-axis joystick. Experimental results were obtained comparing the proposed solution with other controller devices in different scenarios and using a software platform that simulates the kinematics and dynamics of the vehicle. (authors)« less

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

NASA Astrophysics Data System (ADS)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

Propagation of Plasma Bunches through a Transverse Magnetic Barrier

NASA Astrophysics Data System (ADS)

Bishaev, A. M.; Gavrikov, M. B.; Kozintseva, M. V.; Savel'ev, V. V.

2018-01-01

The injection of a plasma bunch into a multipolar trap can be applied to fill the trap with a plasma. The injection of the bunch into a tokamak-like trap can be considered an additional means for controlling the processes of plasma heating and fuel delivery to the central zone of a thermonuclear reactor. In both cases, the bunch is injected normally to the magnetic field of the trap. It has been shown theoretically, experimentally, and by numerical simulation that the depth of plasma bunch penetration into the magnetic field varies in direct proportion to the bunch energy and in inverse proportion to the magnetic pressure and the cross-sectional area of the plasma bunch. The data of this work allow researchers to estimate the values of plasma bunch parameters at which the bunch will be trapped. As a result, the process of plasma bunch trapping has been optimized.

Epidemiology of a thermonuclear bomb-burst over Nashville, Tennessee: a theoretic study

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

Quinn, R.W.

1983-07-01

A thermonuclear bomb explosion over any city in the world would have a devastating effect on the population and environment. For those who survive, with or without injuries, life would become primitive with little or no uncontaminated food or water, and with inadequate housing, fuel, and medical care, resulting in a breakdown of family and interpersonal relationships. This theoretic study of the potential outcome of a thermonuclear bomb-burst over Nashville, Tennessee, discusses epidemiologically the wide range of medical and psychologic effects from the direct trauma of blast and fire, widespread epidemics of otherwise controlled disease, long-term chronic illness, genetic damage,more » and catastrophic environmental havoc.« less

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

NASA Astrophysics Data System (ADS)

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

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

A promising tritium breeding material: Nanostructured 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles

NASA Astrophysics Data System (ADS)

Dang, Chen; Yang, Mao; Gong, Yichao; Feng, Lan; Wang, Hailiang; Shi, Yanli; Shi, Qiwu; Qi, Jianqi; Lu, Tiecheng

2018-03-01

As an advanced tritium breeder material for the fusion reactor blanket of the International Thermonuclear Experimental Reactor (ITER), Li2TiO3-Li4SiO4 biphasic ceramic has attracted widely attention due to its merits. In this paper, the uniform precursor powders were prepared by hydrothermal method, and nanostructured 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles were fabricated by an indirect wet method at the first time. In addition, the composition dependence (x/y) of their microstructure characteristics and mechanical properties were investigated. The results indicated that the crush load of biphasic ceramic pebbles was better than that of single phase ceramic pebbles under identical conditions. The 2Li2TiO3-Li4SiO4 ceramic pebbles have good morphology, small grain size (90 nm), satisfactory crush load (37.8 N) and relative density (81.8 %T.D.), which could be a promising breeding material in the future fusion reactor.

Balanced biomedical program plan. Volume X. Fusion analysis for and environmental research

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

Not Available

1976-06-01

In this draft planning document for health and environmental research needs relevant to the development of fusion technology, an attempt is made to integrate input from the participating laboratories on the basis of the King-Muir study categories. The general description covers only those concepts and features that are considered important to an understanding of possible and probable effects of thermonuclear reactors on health and the environment. Appendixes are included which reflect an understanding of three areas of special interest: materials requirements, effects from magnetic fields, and tritium effects.

Characterization of the ITER model negative ion source during long pulse operation

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

Hemsworth, R.S.; Boilson, D.; Crowley, B.

2006-03-15

It is foreseen to operate the neutral beam system of the International Thermonuclear Experimental Reactor (ITER) for pulse lengths extending up to 1 h. The performance of the KAMABOKO III negative ion source, which is a model of the source designed for ITER, is being studied on the MANTIS test bed at Cadarache. This article reports the latest results from the characterization of the ion source, in particular electron energy distribution measurements and the comparison between positive ion and negative ion extraction from the source.

Matching an Inductive Accumulator and a System of Acceleration of a Liner with Limitation of the Breaking Voltage,

DTIC Science & Technology

In this work calculations are made of the efficiencies of acceleration of a liner from an inductive accumulator in the mode theta-pinch and Z-pinch...to the speed of the liner . Estimations have been made of the necessary power at the moment of switching the current on the basis of considerations of...the stability of the pinch effect of the liner . The level of energies necessary for the creation of a thermonuclear reactor on the basis of theta

Re-weldability tests of irradiated 316L(N) stainless steel using laser welding technique

NASA Astrophysics Data System (ADS)

Yamada, Hirokazu; Kawamura, Hiroshi; Tsuchiya, Kunihiko; Kalinin, George; Kohno, Wataru; Morishima, Yasuo

2002-12-01

SS316L(N)-IG is the candidate material for the in-vessel and ex-vessel components of fusion reactors such as ITER (International Thermonuclear Experimental Reactor). This paper describes a study on re-weldability of un-irradiated and/or irradiated SS316L(N)-IG and the effect of helium generation on the mechanical properties of the weld joint. The laser welding process is used for re-welding of the water cooling branch pipeline repairs. It is clarified that re-welding of SS316L(N)-IG irradiated up to about 0.2 dpa (3.3 appm He) can be carried out without a serious deterioration of tensile properties due to helium accumulation. Therefore, repair of the ITER blanket cooling pipes can be performed by the laser welding process.

Velocity space instabilities of alpha particles in tokamak reactors

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

Sigmar, D.J.

1979-01-01

In this lecture on high frequency instability due to isotropic hollow alpha velocity distributions it was first shown that such distributions can actually arise under thermonuclear conditions in a tokamak reactor, particularly for the case of imperfect alpha particle confinement. The toroidal geometry (i.e., the poloidal variation of the alpha gyrofrequency) then leads to linear instability of the compressional Alfven wave ..omega.. = C/sub A/k/sub perpendicular/ with k/sub parallel/ congruent to O, k/sub perpendicular/ rho/sub ..cap alpha../ greater than or equal to 1, v/sub ..cap alpha../ > C/sub A/, at the low harmonics ..omega.. congruent to n ..omega../sub c..cap alpha../.more » Thus the free energy of the inverted alpha distribution is accessible and produces anomalously rapid diffusion of F/sub ..cap alpha../(v/sub perpendicular/). (MOW)« less

Tritium environmental transport studies at TFTR

NASA Astrophysics Data System (ADS)

Ritter, P. D.; Dolan, T. J.; Longhurst, G. R.

1993-06-01

Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a week after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER).

The choice of the energy embedding law in the design of heavy ionic fusion cylindrical targets

NASA Astrophysics Data System (ADS)

Dolgoleva, GV; Zykova, A. I.

2017-10-01

The paper considers the numerical design of heavy ion fusion (FIHIF) targets, which is one of the branches of controlled thermonuclear fusion (CTF). One of the important tasks in the targets design for controlled thermonuclear fusion is the energy embedding selection whereby it is possible to obtain “burning” (the presence of thermonuclear reactions) of the working DT region. The work is devoted to the rapid ignition of FIHIF targets by means of an additional short-term energy contribution to the DT substance already compressed by massively more longer by energy embedding. This problem has been fairly well studied for laser targets, but this problem is new for heavy ion fusion targets. Maximum momentum increasing is very technically difficult and expensive on modern FIHIF installations. The work shows that the additional energy embedding (“igniting” impulse) reduces the requirements to the maximum impulse. The purpose of this work is to research the ignition impulse effect on the FIHIF target parameters.

Accelerator based fusion reactor

NASA Astrophysics Data System (ADS)

Liu, Keh-Fei; Chao, Alexander Wu

2017-08-01

A feasibility study of fusion reactors based on accelerators is carried out. We consider a novel scheme where a beam from the accelerator hits the target plasma on the resonance of the fusion reaction and establish characteristic criteria for a workable reactor. We consider the reactions d+t\\to n+α,d+{{}3}{{H}\\text{e}}\\to p+α , and p+{{}11}B\\to 3α in this study. The critical temperature of the plasma is determined from overcoming the stopping power of the beam with the fusion energy gain. The needed plasma lifetime is determined from the width of the resonance, the beam velocity and the plasma density. We estimate the critical beam flux by balancing the energy of fusion production against the plasma thermo-energy and the loss due to stopping power for the case of an inert plasma. The product of critical flux and plasma lifetime is independent of plasma density and has a weak dependence on temperature. Even though the critical temperatures for these reactions are lower than those for the thermonuclear reactors, the critical flux is in the range of {{10}22}-{{10}24}~\\text{c}{{\\text{m}}-2}~{{\\text{s}}-1} for the plasma density {ρt}={{10}15}~\\text{c}{{\\text{m}}-3} in the case of an inert plasma. Several approaches to control the growth of the two-stream instability are discussed. We have also considered several scenarios for practical implementation which will require further studies. Finally, we consider the case where the injected beam at the resonance energy maintains the plasma temperature and prolongs its lifetime to reach a steady state. The equations for power balance and particle number conservation are given for this case.

Inertial-confinement fusion with lasers

DOE PAGES

Betti, R.; Hurricane, O. A.

2016-05-03

The quest for controlled fusion energy has been ongoing for over a half century. The demonstration of ignition and energy gain from thermonuclear fuels in the laboratory has been a major goal of fusion research for decades. Thermonuclear ignition is widely considered a milestone in the development of fusion energy, as well as a major scientific achievement with important applications to national security and basic sciences. The U.S. is arguably the world leader in the inertial con fment approach to fusion and has invested in large facilities to pursue it with the objective of establishing the science related to themore » safety and reliability of the stockpile of nuclear weapons. Even though significant progress has been made in recent years, major challenges still remain in the quest for thermonuclear ignition via laser fusion.« less

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

NASA Astrophysics Data System (ADS)

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

2017-10-01

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

Quasi-linear modeling of lower hybrid current drive in ITER and DEMO

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

Cardinali, A., E-mail: alessandro.cardinali@enea.it; Cesario, R.; Panaccione, L.

2015-12-10

First pass absorption of the Lower Hybrid waves in thermonuclear devices like ITER and DEMO is modeled by coupling the ray tracing equations with the quasi-linear evolution of the electron distribution function in 2D velocity space. As usually assumed, the Lower Hybrid Current Drive is not effective in a plasma of a tokamak fusion reactor, owing to the accessibility condition which, depending on the density, restricts the parallel wavenumber to values greater than n{sub ∥crit} and, at the same time, to the high electron temperature that would enhance the wave absorption and then restricts the RF power deposition to themore » very periphery of the plasma column (near the separatrix). In this work, by extensively using the “ray{sup star}” code, a parametric study of the propagation and absorption of the LH wave as function of the coupled wave spectrum (as its width, and peak value), has been performed very accurately. Such a careful investigation aims at controlling the power deposition layer possibly in the external half radius of the plasma, thus providing a valuable aid to the solution of how to control the plasma current profile in a toroidal magnetic configuration, and how to help the suppression of MHD mode that can develop in the outer part of the plasma. This analysis is useful not only for exploring the possibility of profile control of a pulsed operation reactor as well as the tearing mode stabilization, but also in order to reconsider the feasibility of steady state regime for DEMO.« less

The comparison of extraction of energy in two-cascade and one-cascade targets

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

Dolgoleva, G. V., E-mail: dolgg@list.ru; Ponomarev, I. V., E-mail: wingof17@mail.ru

2016-01-15

The paper is devoted to numerical designing of cylindrical microtargets on the basis of shock-free compression. When designing microtargets for the controlled thermonuclear fusion, the core tasks are to select geometry and make-up of layers, and the law of energy embedding as well, which allow receiving of “burning” of deuterium- tritium mix, that is, the existence of thermonuclear reactions of working area. Yet, the energy yield as a result of thermonuclear reactions has to be more than the embedded energy (the coefficient of amplification is more than a unit). So, an important issue is the value of the embedded energy.more » The purpose of the present paper is to study the extraction of energy by working DT area in one-cascade and two-cascade targets. A bigger extraction of energy will contribute to a better burning of DT mix and a bigger energy yield as a result of thermonuclear reactions. The comparison of analytical results to numerical calculations is carried out. The received results show advantages of a two-cascade target compared to a one-cascade one.« less

Review of fusion synfuels

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

Fillo, J.A.

1980-01-01

Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of approx. 40 to 60% and hydrogen production efficiencies by high-temperature electrolysis of approx. 50 to 65% are projected for fusion reactors using high-temperatures blankets. Fusion/coal symbiotic systems appear economically promising for the first generation of commercial fusion synfuels plants. Coal production requirements and the environmental effects of large-scale coal usage would be greatly reduced by a fusion/coal system. In the long term, there could be a gradual transition to an inexhaustible energy system based solely on fusion.

Dielectric strength of irradiated fiber reinforced plastics

NASA Astrophysics Data System (ADS)

Humer, Karl; Weber, Harald W.; Hastik, Ronald; Hauser, Hans; Gerstenberg, Heiko

2001-05-01

The insulation system for the toroidal field model coil of international thermonuclear experimental reactor is a fiber reinforced plastic (FRP) laminate, which consists of a combined Kapton/R-glass-fiber reinforcement tape, vacuum-impregnated with an epoxy DGEBA system. Pure disk-shaped laminates, disk-shaped FRP/stainless-steel sandwiches, and conductor insulation prototypes were irradiated at 5 K in a fission reactor up to a fast neutron fluence of 10 22 m -2 ( E>0.1 MeV) to investigate the radiation induced degradation of the dielectric strength of the insulation system. After warm-up to room temperature, swelling, weight loss, and the breakdown strength were measured at 77 K. The sandwich swells by 4% at a fluence of 5×10 21 m -2 and by 9% at 1×10 22 m -2. The weight loss of the FRP is 2% at 1×10 22 m -2. The dielectric strength remained unchanged over the whole dose range.

Mechanical properties of irradiated beryllium

NASA Astrophysics Data System (ADS)

Beeston, J. M.; Longhurst, G. R.; Wallace, R. S.; Abeln, S. P.

1992-10-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 × 10 25 n/m 2 ( E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.

The status of beryllium technology for fusion

NASA Astrophysics Data System (ADS)

Scaffidi-Argentina, F.; Longhurst, G. R.; Shestakov, V.; Kawamura, H.

2000-12-01

Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented.

Thermonuclear plasma with autocatalytic thermomagnetic current amplification by nuclear reactions from fusion neutrons

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

Winterberg, F.

2006-03-15

It is proposed to use the neutrons released from a deuterium-tritium or deuterium-deuterium fusion reaction to drive thermomagnetic currents in a plasma corona surrounding the fusion plasma through the heating of the corona with nuclear reactions by the neutrons released in the fusion reaction. Because the neutron reaction cross sections are larger for slow neutrons, it is proposed to slow them down in a moderator separated from the hot plasma of the corona, giving the configuration a striking similarity to a heterogeneous nuclear fission reactor. While in a fission reactor the separation makes possible a growing neutron chain reaction, itmore » here makes possible the autocatalytic amplification of the thermomagnetic currents by an increase of the fusion reaction rate through a rise of the plasma pressure by the magnetic pressure of the thermomagnetic currents. This is expected to substantially increase the n{tau} product over its Lawson value.« less

A laser scanning system for metrology and viewing in ITER

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

Spampinato, P.T.; Barry, R.E.; Menon, M.M.

1996-05-01

The construction and operation of a next-generation fusion reactor will require metrology to achieve and verify precise alignment of plasma-facing components and inspection in the reactor vessel. The system must be compatible with the vessel environment of high gamma radiation (10{sup 4} Gy/h), ultra-high-vacuum (10{sup {minus}8} torr), and elevated temperature (200 C). The high radiation requires that the system be remotely deployed. A coherent frequency modulated laser radar-based system will be integrated with a remotely operated deployment mechanism to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics tomore » the laser source and imaging units that are located outside of a biological shield. The deployment mechanism will be a mast-like positioning system. Radiation-damage tests will be conducted on critical sensor components at Oak Ridge National Laboratory to determine threshold damage levels and effects on data transmission. This paper identifies the requirements for International Thermonuclear Experimental Reactor metrology and viewing and describes a remotely operated precision ranging and surface mapping system.« less

Comparison of the recently proposed super-Marx generator approach to thermonuclear ignition with the deuterium-tritium laser fusion-fission hybrid concept by the Lawrence Livermore National Laboratory

DOE PAGES

Winterberg, F.

2009-01-01

The recently proposed super-Marx generator pure deuterium microdetonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser deuterium-tritium fusion-fission hybrid concept (LIFE). In a super-Marx generator, a large number of ordinary Marx generators charge up a much larger second stage ultrahigh voltage Marx generator from which for the ignition of a pure deuterium microexplosion an intense GeV ion beam can be extracted. Typical examples of the LIFE concept are a fusion gain of 30 and a fission gain of 10, making up a total gain of 300, with about ten times more energy released into fissionmore » as compared to fusion. This means the substantial release of fission products, as in fissionless pure fission reactors. In the super-Marx approach for the ignition of pure deuterium microdetonation, a gain of the same magnitude can, in theory, be reached. If feasible, the super-Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of thermonuclear microexplosions.« less

Dynamic simulations for preparing the acceptance test of JT-60SA cryogenic system

NASA Astrophysics Data System (ADS)

Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

2016-12-01

Power generation in the future could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to around 4.5 K. Within this frame, an experimental tokamak device, JT-60SA is currently under construction in Naka (Japan). The plasma works cyclically and the coil system is subject to pulsed heat loads. In order to size the refrigerator close to the average power and hence optimizing investment and operational costs, measures have to be taken to smooth the heat load. Here we present a dynamic model of the JT-60SA's Auxiliary Cold box (ACB) for preparing the acceptance tests of the refrigeration system planned in 2016 in Naka. The aim of this study is to simulate the pulsed load scenarios using different process controls. All the simulations have been performed with EcosimPro® and the associated cryogenic library: CRYOLIB.

Nuclear-Physics Aspects of Controlled Thermonuclear Fusion: Analysis of Promising Fuels and Gamma-Ray Diagnostics of Hot Plasma

NASA Astrophysics Data System (ADS)

Voronchev, V. T.; Kukulin, V. I.

2000-12-01

A brief survey of nuclear-physics aspects of the problems of controlled thermonuclear fusion is given. Attention is paid primarily to choosing and analyzing an optimal composition of a nuclear fuel, reliably extrapolating the cross sections for nuclear reactions to the region of low energies, and exploring gamma-ray methods (as a matter of fact, very promising methods indeed) for diagnostics of hot plasmas (three aspects that are often thought to be the most important ones). In particular, a comparative nuclear-physics analysis of hydrogen, DT, and DD thermonuclear fuels and of their alternatives in the form of D3He, D6Li, DT6Li, H6Li, H11B, and H9Be is performed. Their advantages and disadvantages are highlighted; a spin-polarized fuel is considered; and the current status of nuclear data on the processes of interest is analyzed. A procedure for determining cross sections for nuclear reactions in the deep-subbarrier region is discussed. By considering the example of low-energy D+6Li interactions, it is shown that, at ion temperatures below 100 keV, the inclusion of nuclear-structure factors leads to an additional enhancement of the rate parameters <σv> for the ( d, pt) and ( d, nτ) channels by 10-40%. The possibility of using nuclear reactions that lead to photon emission as a means for determining the ion temperature of a thermonuclear plasma is discussed.

Inertial-confinement fusion with lasers

NASA Astrophysics Data System (ADS)

Betti, R.; Hurricane, O. A.

2016-05-01

The quest for controlled fusion energy has been ongoing for over a half century. The demonstration of ignition and energy gain from thermonuclear fuels in the laboratory has been a major goal of fusion research for decades. Thermonuclear ignition is widely considered a milestone in the development of fusion energy, as well as a major scientific achievement with important applications in national security and basic sciences. The US is arguably the world leader in the inertial confinement approach to fusion and has invested in large facilities to pursue it, with the objective of establishing the science related to the safety and reliability of the stockpile of nuclear weapons. Although significant progress has been made in recent years, major challenges still remain in the quest for thermonuclear ignition via laser fusion. Here, we review the current state of the art in inertial confinement fusion research and describe the underlying physical principles.

Conceptual design of ACB-CP for ITER cryogenic system

NASA Astrophysics Data System (ADS)

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

2012-06-01

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

Remote experimental site concept development

NASA Astrophysics Data System (ADS)

Casper, Thomas A.; Meyer, William; Butner, David

1995-01-01

Scientific research is now often conducted on large and expensive experiments that utilize collaborative efforts on a national or international scale to explore physics and engineering issues. This is particularly true for the current US magnetic fusion energy program where collaboration on existing facilities has increased in importance and will form the basis for future efforts. As fusion energy research approaches reactor conditions, the trend is towards fewer large and expensive experimental facilities, leaving many major institutions without local experiments. Since the expertise of various groups is a valuable resource, it is important to integrate these teams into an overall scientific program. To sustain continued involvement in experiments, scientists are now often required to travel frequently, or to move their families, to the new large facilities. This problem is common to many other different fields of scientific research. The next-generation tokamaks, such as the Tokamak Physics Experiment (TPX) or the International Thermonuclear Experimental Reactor (ITER), will operate in steady-state or long pulse mode and produce fluxes of fusion reaction products sufficient to activate the surrounding structures. As a direct consequence, remote operation requiring robotics and video monitoring will become necessary, with only brief and limited access to the vessel area allowed. Even the on-site control room, data acquisition facilities, and work areas will be remotely located from the experiment, isolated by large biological barriers, and connected with fiber-optics. Current planning for the ITER experiment includes a network of control room facilities to be located in the countries of the four major international partners; USA, Russian Federation, Japan, and the European Community.

Nuclear transmutation in steels

NASA Astrophysics Data System (ADS)

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

2009-05-01

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

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

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

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

Development of a cryogenic load frame for the neutron diffractometer at Takumi in Japan Proton Accelerator Research Complex

NASA Astrophysics Data System (ADS)

Jin, Xinzhe; Nakamoto, Tatsushi; Harjo, Stefanus; Hemmi, Tsutomu; Umeno, Takahiro; Ogitsu, Toru; Yamamoto, Akira; Sugano, Michinaka; Aizawa, Kazuya; Abe, Jun; Gong, Wu; Iwahashi, Takaaki

2013-06-01

To prepare for projects such as the Large Hadron Collider upgrade, International Thermonuclear Experimental Reactor and Demonstration reactor, it is important to form a clear understanding of stress-strain properties of the materials that make up superconducting magnets. Thus, we have been studying the mechanical properties of superconducting wires using neutron diffraction measurements. To simulate operational conditions such as temperature, stress, and strain, we developed a cryogenic load frame for stress-strain measurements of materials using a neutron diffractometer at Japan Proton Accelerator Research Complex (J-PARC) Takumi beam line. The maximum load that can be applied to a sample using an external driving machine is 50 kN. Using a Gifford-MacMahon cryocooler, samples can be measured down to temperatures below 10 K when loaded. In the present paper, we describe the details of the cryogenic load frame with its test results by using type-304 stainless steel wire.

Development of a cryogenic load frame for the neutron diffractometer at Takumi in Japan Proton Accelerator Research Complex

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

Jin, Xinzhe; Nakamoto, Tatsushi; Ogitsu, Toru

2013-06-15

To prepare for projects such as the Large Hadron Collider upgrade, International Thermonuclear Experimental Reactor and Demonstration reactor, it is important to form a clear understanding of stress-strain properties of the materials that make up superconducting magnets. Thus, we have been studying the mechanical properties of superconducting wires using neutron diffraction measurements. To simulate operational conditions such as temperature, stress, and strain, we developed a cryogenic load frame for stress-strain measurements of materials using a neutron diffractometer at Japan Proton Accelerator Research Complex (J-PARC) Takumi beam line. The maximum load that can be applied to a sample using an externalmore » driving machine is 50 kN. Using a Gifford-MacMahon cryocooler, samples can be measured down to temperatures below 10 K when loaded. In the present paper, we describe the details of the cryogenic load frame with its test results by using type-304 stainless steel wire.« less

Development of a cryogenic load frame for the neutron diffractometer at Takumi in Japan Proton Accelerator Research Complex.

PubMed

Jin, Xinzhe; Nakamoto, Tatsushi; Harjo, Stefanus; Hemmi, Tsutomu; Umeno, Takahiro; Ogitsu, Toru; Yamamoto, Akira; Sugano, Michinaka; Aizawa, Kazuya; Abe, Jun; Gong, Wu; Iwahashi, Takaaki

2013-06-01

To prepare for projects such as the Large Hadron Collider upgrade, International Thermonuclear Experimental Reactor and Demonstration reactor, it is important to form a clear understanding of stress-strain properties of the materials that make up superconducting magnets. Thus, we have been studying the mechanical properties of superconducting wires using neutron diffraction measurements. To simulate operational conditions such as temperature, stress, and strain, we developed a cryogenic load frame for stress-strain measurements of materials using a neutron diffractometer at Japan Proton Accelerator Research Complex (J-PARC) Takumi beam line. The maximum load that can be applied to a sample using an external driving machine is 50 kN. Using a Gifford-MacMahon cryocooler, samples can be measured down to temperatures below 10 K when loaded. In the present paper, we describe the details of the cryogenic load frame with its test results by using type-304 stainless steel wire.

Magnetless magnetic fusion

NASA Astrophysics Data System (ADS)

Beklemishev, A. D.; Tajima, T.

1994-02-01

The authors propose a concept of thermonuclear fusion reactor in which the plasma pressure is balanced by direct gas-wall interaction in a high-pressure vessel. The energy confinement is achieved by means of the self-contained toroidal magnetic configuration sustained by an external current drive or charged fusion products. This field structure causes the plasma pressure to decrease toward the inside of the discharge and thus it should be magnetohydrodynamically stable. The maximum size, temperature and density profiles of the reactor are estimated. An important feature of confinement physics is the thin layer of cold gas at the wall and the adjacent transitional region of dense arc-like plasma. The burning condition is determined by the balance between these nonmagnetized layers and the current-carrying plasma. They suggest several questions for future investigation, such as the thermal stability of the transition layer and the possibility of an effective heating and current drive behind the dense edge plasma. The main advantage of this scheme is the absence of strong external magnets and, consequently, potentially cheaper design and lower energy consumption.

Comparison of the Recently proposed Super Marx Generator Approach to Thermonuclear Ignition with the DT Laser Fusion-Fission Hybrid Concept (LIFE) by the Lawrence Livermore National Laboratory.

NASA Astrophysics Data System (ADS)

Winterberg, Friedwardt

2009-05-01

The recently proposed Super Marx pure deuterium micro-detonation ignition concept [1] is compared to the Lawrence Livermore National Ignition Facility (NIF) laser DT fusion-fission hybrid concept (LIFE) [2]. A typical example of the LIFE concept is a fusion gain 30, and a fission gain of 10, making up for a total gain of 300, with about 10 times more energy released into fission as compared to fusion. This means a substantial release of fission products, as in fusion-less pure fission reactors. In the Super Marx approach for the ignition of a pure deuterium micro-detonation gains of the same magnitude can in theory be reached. If the theoretical prediction can be supported by more elaborate calculations, the Super Marx approach is likely to make lasers obsolete as a means for the ignition of thermonuclear micro-explosions. [1] ``Ignition of a Deuterium Micro-Detonation with a Gigavolt Super Marx Generator,'' Winterberg, F., Journal of Fusion Energy, Springer, 2008. http://www.springerlink.com/content/r2j046177j331241/fulltext.pdf. [2] ``LIFE: Clean Energy from Nuclear Waste,'' https://lasers.llnl.gov/missions/energy&_slash;for&_slash;the&_slash;future/life/

Fusion power for space propulsion.

NASA Technical Reports Server (NTRS)

Roth, R.; Rayle, W.; Reinmann, J.

1972-01-01

Principles of operation, interplanetary orbit-to-orbit mission capabilities, technical problems, and environmental safeguards are examined for thermonuclear fusion propulsion systems. Two systems examined include (1) a fusion-electric concept in which kinetic energy of charged particles from the plasma is converted into electric power (for accelerating the propellant in an electrostatic thrustor) by the van de Graaf generator principle and (2) the direct fusion rocket in which energetic plasma lost from the reactor has a suitable amount of added propellant to obtain the optimum exhaust velocity. The deuterium-tritium and the deuterium/helium-3 reactions are considered as suitable candidates, and attention is given to problems of cryogenic refrigeration systems, magnet shielding, and high-energy particle extraction and guidance.

Linus cycle calculations including plasma transport and resistive flux loss in an incompressible liner

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

Quimby, D.C.; Hoffman, A.L.; Vlases, G.C.

1980-08-01

In the LINUS fusion reactor concept, a rotating liquid metal liner is used for reversible mechanical compression of thermonuclear plasmas, where a vacuum field buffer zone is used between the plasma and wall to reduce transport losses. A one-dimensional plasma transport and burn code, including incompressible liner dynamics with heat transfer and temperature dependent flux diffusion in the liquid metal, is used to model LINUS cycles. The effects of compressibility are treated as a perturbation. Numerical coefficients are derived for simple LINUS scaling laws. The particular case of plasma contact with the liquid metal is studied to determine the effectmore » on LINUS performance.« less

Formation and sustainment of internal transport barriers in the International Thermonuclear Experimental Reactor with the baseline heating mixa)

NASA Astrophysics Data System (ADS)

Poli, Francesca M.; Kessel, Charles E.

2013-05-01

Plasmas with internal transport barriers (ITBs) are a potential and attractive route to steady-state operation in ITER. These plasmas exhibit radially localized regions of improved confinement with steep pressure gradients in the plasma core, which drive large bootstrap current and generate hollow current profiles and negative magnetic shear. This work examines the formation and sustainment of ITBs in ITER with electron cyclotron heating and current drive. The time-dependent transport simulations indicate that, with a trade-off of the power delivered to the equatorial and to the upper launcher, the sustainment of steady-state ITBs can be demonstrated in ITER with the baseline heating configuration.

ITER Magnet Feeder: Design, Manufacturing and Integration

NASA Astrophysics Data System (ADS)

CHEN, Yonghua; ILIN, Y.; M., SU; C., NICHOLAS; BAUER, P.; JAROMIR, F.; LU, Kun; CHENG, Yong; SONG, Yuntao; LIU, Chen; HUANG, Xiongyi; ZHOU, Tingzhi; SHEN, Guang; WANG, Zhongwei; FENG, Hansheng; SHEN, Junsong

2015-03-01

The International Thermonuclear Experimental Reactor (ITER) feeder procurement is now well underway. The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) in the last 2 years along with analyses and qualification activities. The feeder design is being progressively finalized. In addition, the preparation of qualification and manufacturing are well scheduled at ASIPP. This paper mainly presents the design, the overview of manufacturing and the status of integration on the ITER magnet feeders. supported by the National Special Support for R&D on Science and Technology for ITER (Ministry of Public Security of the People's Republic of China-MPS) (No. 2008GB102000)

Di(cyanate Ester) Networks Based on Alternative Fluorinated Bisphenols with Extremely Low Water Uptake

DTIC Science & Technology

2013-11-01

radomes6  to magnet casings  for  thermonuclear  fusion  reactors7,8  and  support  structures   for  interplanetary  space  probes.9    These  properties...because  the  conversion  is  relatively  easy  to  quantify via methods such as FT‐IR spectroscopy,17 investiga‐ tions  of  structure ‐property...the  high‐temperature  resistance  and  mechanical  stiffness  associated with aromatic content, which  limits  their poten‐ tial  for high

Hydrogen transport behavior of beryllium

NASA Astrophysics Data System (ADS)

Anderl, R. A.; Hankins, M. R.; Longhurst, G. R.; Pawelko, R. J.; Macaulay-Newcombe, R. G.

1992-12-01

Beryllium is being evaluated for use as a plasma-facing material in the International Thermonuclear Experimental Reactor (ITER). One concern in the evaluation is the retention and permeation of tritium implanted into the plasma-facing surface. We performed laboratory-scale studies to investigate mechanisms that influence hydrogen transport and retention in beryllium foil specimens of rolled powder metallurgy product and rolled ingot cast beryllium. Specimen characterization was accomplished using scanning electron microscopy, Auger electron spectroscopy, and Rutherford backscattering spectrometry (RBS) techniques. Hydrogen transport was investigated using ion-beam permeation experiments and nuclear reaction analysis (NRA). Results indicate that trapping plays a significant role in permeation, re-emission, and retention, and that surface processes at both upstream and downstream surfaces are also important.

Optimization of hybrid power system composed of SMES and flywheel MG for large pulsed load

NASA Astrophysics Data System (ADS)

Niiyama, K.; Yagai, T.; Tsuda, M.; Hamajima, T.

2008-09-01

A superconducting magnetic storage system (SMES) has some advantages such as rapid large power response and high storage efficiency which are superior to other energy storage systems. A flywheel motor generator (FWMG) has large scaled capacity and high reliability, and hence is broadly utilized for a large pulsed load, while it has comparatively low storage efficiency due to high mechanical loss compared with SMES. A fusion power plant such as International Thermo-Nuclear Experimental Reactor (ITER) requires a large and long pulsed load which causes a frequency deviation in a utility power system. In order to keep the frequency within an allowable deviation, we propose a hybrid power system for the pulsed load, which equips the SMES and the FWMG with the utility power system. We evaluate installation cost and frequency control performance of three power systems combined with energy storage devices; (i) SMES with the utility power, (ii) FWMG with the utility power, (iii) both SMES and FWMG with the utility power. The first power system has excellent frequency power control performance but its installation cost is high. The second system has inferior frequency control performance but its installation cost is the lowest. The third system has good frequency control performance and its installation cost is attained lower than the first power system by adjusting the ratio between SMES and FWMG.

Long-pulse beam acceleration of MeV-class H(-) ion beams for ITER NB accelerator.

PubMed

Umeda, N; Kashiwagi, M; Taniguchi, M; Tobari, H; Watanabe, K; Dairaku, M; Yamanaka, H; Inoue, T; Kojima, A; Hanada, M

2014-02-01

In order to realize neutral beam systems in International Thermonuclear Experimental Reactor whose target is to produce a 1 MeV, 200 A/m(2) during 3600 s D(-) ion beam, the electrostatic five-stages negative ion accelerator so-called "MeV accelerator" has been developed at Japan Atomic Energy Agency. To extend pulse length, heat load of the acceleration grids was reduced by controlling the ion beam trajectory. Namely, the beam deflection due to the residual magnetic field of filter magnet was suppressed with the newly developed extractor with a 0.5 mm off-set aperture displacement. The new extractor improved the deflection angle from 6 mrad to 1 mrad, resulting in the reduction of direct interception of negative ions from 23% to 15% of the total acceleration power, respectively. As a result, the pulse length of 130 A/m(2), 881 keV H(-) ion beam has been successfully extended from a previous value of 0.4 s to 8.7 s. This is the first long pulse negative ion beam acceleration over 100 MW/m(2).

Role of the lower hybrid spectrum in the current drive modeling for DEMO scenarios

NASA Astrophysics Data System (ADS)

Cardinali, A.; Castaldo, C.; Cesario, R.; Santini, F.; Amicucci, L.; Ceccuzzi, S.; Galli, A.; Mirizzi, F.; Napoli, F.; Panaccione, L.; Schettini, G.; Tuccillo, A. A.

2017-07-01

The active control of the radial current density profile is one of the major issues of thermonuclear fusion energy research based on magnetic confinement. The lower hybrid current drive could in principle be used as an efficient tool. However, previous understanding considered the electron temperature envisaged in a reactor at the plasma periphery too large to allow penetration of the coupled radio frequency (RF) power due to strong Landau damping. In this work, we present new numerical results based on quasilinear theory, showing that the injection of power spectra with different {n}// widths of the main lobe produce an RF-driven current density profile spanning most of the outer radial half of the plasma ({n}// is the refractive index in a parallel direction to the confinement magnetic field). Plasma kinetic profiles envisaged for the DEMO reactor are used as references. We demonstrate the robustness of the modeling results concerning the key role of the spectral width in determining the lower hybrid-driven current density profile. Scans of plasma parameters are extensively carried out with the aim of excluding the possibility that any artefact of the utilised numerical modeling would produce any novelty. We neglect here the parasitic effect of spectral broadening produced by linear scattering due to plasma density fluctuations, which mainly occurs for low magnetic field devices. This effect will be analyzed in other work that completes the report on the present breakthrough.

Plasma cleaning of ITER first mirrors

NASA Astrophysics Data System (ADS)

Moser, L.; Marot, L.; Steiner, R.; Reichle, R.; Leipold, F.; Vorpahl, C.; Le Guern, F.; Walach, U.; Alberti, S.; Furno, I.; Yan, R.; Peng, J.; Ben Yaala, M.; Meyer, E.

2017-12-01

Nuclear fusion is an extremely attractive option for future generations to compete with the strong increase in energy consumption. Proper control of the fusion plasma is mandatory to reach the ambitious objectives set while preserving the machine’s integrity, which requests a large number of plasma diagnostic systems. Due to the large neutron flux expected in the International Thermonuclear Experimental Reactor (ITER), regular windows or fibre optics are unusable and were replaced by so-called metallic first mirrors (FMs) embedded in the neutron shielding, forming an optical labyrinth. Materials eroded from the first wall reactor through physical or chemical sputtering will migrate and will be deposited onto mirrors. Mirrors subject to net deposition will suffer from reflectivity losses due to the deposition of impurities. Cleaning systems of metallic FMs are required in more than 20 optical diagnostic systems in ITER. Plasma cleaning using radio frequency (RF) generated plasmas is currently being considered the most promising in situ cleaning technique. An update of recent results obtained with this technique will be presented. These include the demonstration of cleaning of several deposit types (beryllium, tungsten and beryllium proxy, i.e. aluminium) at 13.56 or 60 MHz as well as large scale cleaning (mirror size: 200 × 300 mm2). Tests under a strong magnetic field up to 3.5 T in laboratory and first experiments of RF plasma cleaning in EAST tokamak will also be discussed. A specific focus will be given on repetitive cleaning experiments performed on several FM material candidates.

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

DOE PAGES

Okabayashi, M.; Zanca, P.; Strait, E. J.; ...

2016-11-25

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

NASA Astrophysics Data System (ADS)

Okabayashi, M.; Zanca, P.; Strait, E. J.; Garofalo, A. M.; Hanson, J. M.; In, Y.; La Haye, R. J.; Marrelli, L.; Martin, P.; Paccagnella, R.; Paz-Soldan, C.; Piovesan, P.; Piron, C.; Piron, L.; Shiraki, D.; Volpe, F. A.; DIII-D, The; RFX-mod Teams

2017-01-01

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. Here we have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance is provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. Here β N is defined as β N  =  β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. The internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.

Summary Report on Controlled Thermonuclear Synthesis,

DTIC Science & Technology

stellarator and the tokamak. Adiabatic magnetic traps are also briefly discussed, as well as the plasma focus . The paper is a very brief generalization of the current state of high-temperature plasma physics. (Author)

Lawson criterion in cyclotron heating

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

Demutskii, V.P.; Polovin, R.V.

1975-07-01

Stochastic heating of plasma particles is of great interest for controlled thermonuclear reactions. The ion velocity distribution function is described for the case of cyclotron heating. The Lawson criterion applied to this distribution is described. (MOW)

Simulation of plasma-surface interactions in a fusion reactor by means of QSPA plasma streams: recent results and prospects

NASA Astrophysics Data System (ADS)

Garkusha, I. E.; Aksenov, N. N.; Byrka, O. V.; Makhlaj, V. A.; Herashchenko, S. S.; Malykhin, S. V.; Petrov, Yu V.; Staltsov, V. V.; Surovitskiy, S. V.; Wirtz, M.; Linke, J.; Sadowski, M. J.; Skladnik-Sadowska, E.

2016-09-01

This paper is devoted to plasma-surface interaction issues at high heat-loads which are typical for fusion reactors. For the International Thermonuclear Experimental Reactor (ITER), which is now under construction, the knowledge of erosion processes and the behaviour of various constructional materials under extreme conditions is a very critical issue, which will determine a successful realization of the project. The most important plasma-surface interaction (PSI) effects in 3D geometry have been studied using a QSPA Kh-50 powerful quasi-stationary plasma accelerator. Mechanisms of the droplet and dust generation have been investigated in detail. It was found that the droplets emission from castellated surfaces has a threshold character and a cyclic nature. It begins only after a certain number of the irradiating plasma pulses when molten and shifted material is accumulated at the edges of the castellated structure. This new erosion mechanism, connected with the edge effects, results in an increase in the size of the emitted droplets (as compared with those emitted from a flat surface). This mechanism can even induce the ejection of sub-mm particles. A concept of a new-generation QSPA facility, the current status of this device maintenance, and prospects for further experiments are also presented.

Normal operation and maintenance safety lessons from the ITER US PbLi test blanket module program for a US FNSF and DEMO

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

L. C. Cadwallader; C. P. C. Wong; M. Abdou

2014-10-01

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module andmore » blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the International Thermonuclear Experimental Reactor (ITER) Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.« less

Extruder system for high-throughput/steady-state hydrogen ice supply and application for pellet fueling of reactor-scale fusion experiments

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

Combs, S.K.; Foust, C.R.; Qualls, A.L.

Pellet injection systems for the next-generation fusion devices, such as the proposed International Thermonuclear Experimental Reactor (ITER), will require feed systems capable of providing a continuous supply of hydrogen ice at high throughputs. A straightforward concept in which multiple extruder units operate in tandem has been under development at the Oak Ridge National Laboratory. A prototype with three large-volume extruder units has been fabricated and tested in the laboratory. In experiments, it was found that each extruder could provide volumetric ice flow rates of up to {approximately}1.3 cm{sup 3}/s (for {approximately}10 s), which is sufficient for fueling fusion reactors atmore » the gigawatt power level. With the three extruders of the prototype operating in sequence, a steady rate of {approximately}0.33 cm{sup 3}/s was maintained for a duration of 1 h. Even steady-state rates approaching the full ITER design value ({approximately}1 cm{sup 3}/s) may be feasible with the prototype. However, additional extruder units (1{endash}3) would facilitate operations at the higher throughputs and reduce the duty cycle of each unit. The prototype can easily accommodate steady-state pellet fueling of present large tokamaks or other near-term plasma experiments.« less

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Fisch, Nathaniel J.

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to establish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated in the plasma.

System and method for generating steady state confining current for a toroidal plasma fusion reactor

DOEpatents

Bers, Abraham

1981-01-01

A system for generating steady state confining current for a toroidal plasma fusion reactor providing steady-state generation of the thermonuclear power. A dense, hot toroidal plasma is initially prepared with a confining magnetic field with toroidal and poloidal components. Continuous wave RF energy is injected into said plasma to estalish a spectrum of traveling waves in the plasma, where the traveling waves have momentum components substantially either all parallel, or all anti-parallel to the confining magnetic field. The injected RF energy is phased to couple to said traveling waves with both a phase velocity component and a wave momentum component in the direction of the plasma traveling wave components. The injected RF energy has a predetermined spectrum selected so that said traveling waves couple to plasma electrons having velocities in a predetermined range .DELTA.. The velocities in the range are substantially greater than the thermal electron velocity of the plasma. In addition, the range is sufficiently broad to produce a raised plateau having width .DELTA. in the plasma electron velocity distribution so that the plateau electrons provide steady-state current to generate a poloidal magnetic field component sufficient for confining the plasma. In steady state operation of the fusion reactor, the fusion power density in the plasma exceeds the power dissipated inthe plasma.

Near-infrared spectroscopy for burning plasma diagnostic applications.

PubMed

Soukhanovskii, V A

2008-10-01

Ultraviolet and visible (UV-VIS, 200-750 nm) atomic spectroscopy of neutral and ionized fuel species (H, D, T, and Li) and impurities (e.g., He, Be, C, and W) is a key element of plasma control and diagnosis on International Thermonuclear Experimental Reactor and future magnetically confined burning plasma experiments (BPXs). Spectroscopic diagnostic implementation and performance issues that arise in the BPX harsh nuclear environment in the UV-VIS range, e.g., degradation of first mirror reflectivity under charge-exchange atom bombardment (erosion) and impurity deposition, permanent and dynamic loss of window, and optical fiber transmission under intense neutron and gamma-ray fluxes, are either absent or not as severe in the near-infrared (NIR, 750-2000 nm) range. An initial survey of NIR diagnostic applications has been undertaken on the National Spherical Torus Experiment. It is demonstrated that NIR spectroscopy can be used for machine protection and plasma control applications, as well as contribute to plasma performance evaluation and physics studies. Emission intensity estimates demonstrate that NIR measurements are possible in the BPX plasma operating parameter range. Complications in the NIR range due to the parasitic background emissions are expected to occur at very high plasma densities, low impurity densities, and at high plasma-facing component temperatures.

U. S. fusion programs: Struggling to stay in the game

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

Crawford, M.

Funding for the US fusion energy program has suffered and will probably continue to suffer major cuts. A committee hand-picked by Energy Secretary James Watkins urged the Department of Energy to mount an aggressive program to develop fusion power, but congress cut funding from $323 million in 1990 to $275 million in 1991. This portends dire conditions for fusion research and development. Projects to receive top priority are concerned with the tokamaks and to keep the next big machine, the Burning Plasma Experiment, scheduled for beginning of construction in 1993 on schedule. Secretary Watkins is said to want to keepmore » the International Thermonuclear Energy Reactor (ITER) on schedule. ITER would follow the Burning Plasma Experiment.« less

Thermal modeling of W rod armor.

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

Nygren, Richard Einar

2004-09-01

Sandia has developed and tested mockups armored with W rods over the last decade and pioneered the initial development of W rod armor for International Thermonuclear Experimental Reactor (ITER) in the 1990's. We have also developed 2D and 3D thermal and stress models of W rod-armored plasma facing components (PFCs) and test mockups and are applying the models to both short pulses, i.e. edge localized modes (ELMs), and thermal performance in steady state for applications in C-MOD, DiMES testing and ITER. This paper briefly describes the 2D and 3D models and their applications with emphasis on modeling for an ongoingmore » test program that simulates repeated heat loads from ITER ELMs.« less

Final Report on ITER Task Agreement 81-10

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

Brad J. Merrill

An International Thermonuclear Experimental Reactor (ITER) Implementing Task Agreement (ITA) on Magnet Safety was established between the ITER International Organization (IO) and the Idaho National Laboratory (INL) Fusion Safety Program (FSP) during calendar year 2004. The objectives of this ITA were to add new capabilities to the MAGARC code and to use this updated version of MAGARC to analyze unmitigated superconductor quench events for both poloidal field (PF) and toroidal field (TF) coils of the ITER design. This report documents the completion of the work scope for this ITA. Based on the results obtained for this ITA, an unmitigated quenchmore » event in an ITER larger PF coil does not appear to be as severe an accident as in an ITER TF coil.« less

A real-time n/γ digital pulse shape discriminator based on FPGA.

PubMed

Li, Shiping; Xu, Xiufeng; Cao, Hongrui; Yuan, Guoliang; Yang, Qingwei; Yin, Zejie

2013-02-01

A FPGA-based real-time digital pulse shape discriminator has been employed to distinguish between neutrons (n) and gammas (γ) in the Neutron Flux Monitor (NFM) for International Thermonuclear Experimental Reactor (ITER). The discriminator takes advantages of the Field Programmable Gate Array (FPGA) parallel and pipeline process capabilities to carry out the real-time sifting of neutrons in n/γ mixed radiation fields, and uses the rise time and amplitude inspection techniques simultaneously as the discrimination algorithm to observe good n/γ separation. Some experimental results have been presented which show that this discriminator can realize the anticipated goals of NFM perfectly with its excellent discrimination quality and zero dead time. Copyright © 2012 Elsevier Ltd. All rights reserved.

NEUTRON SOURCE USING MAGNETIC COMPRESSION OF PLASMA

DOEpatents

Quinn, W.E.; Elmore, W.C.; Little, E.M.; Boyer, K.; Tuck, J.L.

1961-10-31

A fusion reactor is described that utilizes compression and heating of an ionized thermonuclear fuel by an externally applied magnetic field, thus avoiding reliance on the pinch effect and its associated instability problems. The device consists of a gas-confining ceramic container surrounded by a single circumferential coil having a shape such as to produce a magnetic mirror geometry. A sinusoidally-oscillating, exponentially-damped current is passed circumferentially around the container, through the coil, inducing a circumferential current in the gas. Maximum compression and plasma temperature are obtained at the peak of the current oscillations, coinciding with maximum magnetic field intensity. Enhanced temperatures are obtained in the second and succeeding half cycles because the thermal energy accumulates from one half cycle to the next. (AEC)

High-Q plasmas in the TFTR tokamak

NASA Astrophysics Data System (ADS)

Jassby, D. L.; Barnes, C. W.; Bell, M. G.; Bitter, M.; Boivin, R.; Bretz, N. L.; Budny, R. V.; Bush, C. E.; Dylla, H. F.; Efthimion, P. C.; Fredrickson, E. D.; Hawryluk, R. J.; Hill, K. W.; Hosea, J.; Hsuan, H.; Janos, A. C.; Jobes, F. C.; Johnson, D. W.; Johnson, L. C.; Kamperschroer, J.; Kieras-Phillips, C.; Kilpatrick, S. J.; LaMarche, P. H.; LeBlanc, B.; Mansfield, D. K.; Marmar, E. S.; McCune, D. C.; McGuire, K. M.; Meade, D. M.; Medley, S. S.; Mikkelsen, D. R.; Mueller, D.; Owens, D. K.; Park, H. K.; Paul, S. F.; Pitcher, S.; Ramsey, A. T.; Redi, M. H.; Sabbagh, S. A.; Scott, S. D.; Snipes, J.; Stevens, J.; Strachan, J. D.; Stratton, B. C.; Synakowski, E. J.; Taylor, G.; Terry, J. L.; Timberlake, J. R.; Towner, H. H.; Ulrickson, M.; von Goeler, S.; Wieland, R. M.; Williams, M.; Wilson, J. R.; Wong, K.-L.; Young, K. M.; Zarnstorff, M. C.; Zweben, S. J.

1991-08-01

In the Tokamak Fusion Test Reactor (TFTR) [Plasma Phys. Controlled Fusion 26, 11 (1984)], the highest neutron source strength Sn and D-D fusion power gain QDD are realized in the neutral-beam-fueled and heated ``supershot'' regime that occurs after extensive wall conditioning to minimize recycling. For the best supershots, Sn increases approximately as P1.8b. The highest-Q shots are characterized by high Te (up to 12 keV), Ti (up to 34 keV), and stored energy (up to 4.7 MJ), highly peaked density profiles, broad Te profiles, and lower Zeff. Replacement of critical areas of the graphite limiter tiles with carbon-fiber composite tiles and improved alignment with the plasma have mitigated the ``carbon bloom.'' Wall conditioning by lithium pellet injection prior to the beam pulse reduces carbon influx and particle recycling. Empirically, QDD increases with decreasing pre-injection carbon radiation, and increases strongly with density peakedness [ne(0)/] during the beam pulse. To date, the best fusion results are Sn=5×1016 n/sec, QDD=1.85×10-3, and neutron yield=4.0×1016 n/pulse, obtained at Ip=1.6-1.9 MA and beam energy Eb=95-103 keV, with nearly balanced co- and counter-injected beam power. Computer simulations of supershot plasmas show that typically 50%-60% of Sn arises from beam-target reactions, with the remainder divided between beam-beam and thermonuclear reactions, the thermonuclear fraction increasing with Pb. The simulations predict that QDT=0.3-0.4 would be obtained for the best present plasma conditions, if half the deuterium neutral beams were to be replaced by tritium beams. Somewhat higher values are calculated if D beams are injected into a predominantly tritium target plasma. The projected central beta of fusion alphas is 0.4%-0.6%, a level sufficient for the study of alpha-induced collective effects.

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

An evaluation of multilayer mirrors for the soft x ray and extreme ultraviolet wavelength range that were irradiated with neutrons

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

Regan, S.P.; May, M.J.; Soukhanovskii, V.

1997-01-01

The Plasma Spectroscopy Group at the Johns Hopkins University develops high photon throughput multilayer mirror (MLM) based soft x ray and extreme ultraviolet (XUV 10 {Angstrom}{lt}{lambda}{lt}304 {Angstrom}) spectroscopic diagnostics for magnetically confined fusion plasmas. The D-T reactions in large fusion reactor type devices such as the International Thermonuclear Experimental Reactor will produce neutrons at a rate as high as 5{times}10{sup 19} ns{sup -1}. The MLMs, which are used as dispersive and focusing optics, will not be shielded from these neutrons. In an effort to assess the potential radiation damage, four MLMs (No. 1: Mo/Si, d=87.8 {Angstrom}, Zerodur substrate with 50more » cm concave spherical curvature; No. 2: W/B{sub 4}C, d=22.75 {Angstrom}, Si wafer substrate; No. 3: W/C, d=25.3 {Angstrom}, Si wafer substrate; and No. 4: Mo/Si, d=186.6 {Angstrom}, Si wafer substrate) were irradiated with fast neutrons at the Los Alamos Spallation Radiation Effects Facility (LASREF). The neutron beam at LASREF has an energy distribution that peaks at 1{endash}2 MeV with a tail that extends out to 100 MeV. The MLMs were irradiated to a fast neutron fluence of 1.1{times}10{sup 19} ncm{sup {minus}2} at 270{endash}300{degree}C. A comparison between the dispersive and reflective characteristics of the irradiated MLMs and the corresponding qualities of control samples will be given. {copyright} {ital 1997 American Institute of Physics.}« less

Laser induced sonofusion: A new road toward thermonuclear reactions

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

Sadighi-Bonabi, Rasoul, E-mail: Sadighi@sharif.ir; Gheshlaghi, Maryam; Laser and optics research school, Nuclear Science and Technology Research Institute

2016-03-15

The Possibility of the laser assisted sonofusion is studied via single bubble sonoluminescence (SBSL) in Deuterated acetone (C{sub 3}D{sub 6}O) using quasi-adiabatic and hydro-chemical simulations at the ambient temperatures of 0 and −28.5 °C. The interior temperature of the produced bubbles in Deuterated acetone is 1.6 × 10{sup 6} K in hydro-chemical model and it is reached up to 1.9 × 10{sup 6} K in the laser induced SBSL bubbles. Under these circumstances, temperature up to 10{sup 7} K can be produced in the center of the bubble in which the thermonuclear D-D fusion reactions are promising under the controlled conditions.

Preliminary Process Design of ITER ELM Coil Bracket Brazing

NASA Astrophysics Data System (ADS)

LI, Xiangbin; SHI, Yi

2015-03-01

With the technical requirement of the International Thermonuclear Experimental Reactor (ITER) project, the manufacture and assembly technology of the mid Edge Localized Modes (ELM) coil was developed by the Institute of Plasma Physics, Chinese Academy of Science (ASIPP). As the gap between the bracket and the Stainless Steel jacketed and Mineral Insulated Conductor (SSMIC) can be larger than 0.5 mm instead of 0.01 mm to 0.1 mm as in normal industrial cases, the process of mid ELM coil bracket brazing to the SSMICT becomes quiet challenging, from a technical viewpoint. This paper described the preliminary design of ELM coil bracket brazing to the SSMIC process, the optimal bracket brazing curve and the thermal simulation of the bracket furnace brazing method developed by ANSYS. BAg-6 foil (Bag50Cu34Zn16) plus BAg-1a paste (Bag45CuZnCd) solders were chosen as the brazing filler. By testing an SSMICT prototype, it is shown that the average gap between the bracket and the SSMIC could be controlled to 0.2-0.3 mm, and that there were few voids in the brazing surface. The results also verified that the preliminary design had a favorable heat conducting performance in the bracket.

Sandia technology: Engineering and science applications

NASA Astrophysics Data System (ADS)

Maydew, M. C.; Parrot, H.; Dale, B. C.; Floyd, H. L.; Leonard, J. A.; Parrot, L.

1990-12-01

This report discusses: protecting environment, safety, and health; Sandia's quality initiative; Sandia vigorously pursues technology transfer; scientific and technical education support programs; nuclear weapons development; recognizing battlefield targets with trained artificial neural networks; battlefield robotics: warfare at a distance; a spinning shell sizes up the enemy; thwarting would-be nuclear terrorists; unattended video surveillance system for nuclear facilities; making the skies safer for travelers; onboard instrumentation system to evaluate performance of stockpile bombs; keeping track with lasers; extended-life lithium batteries; a remote digital video link acquires images securely; guiding high-performance missiles with laser gyroscopes; nonvolatile memory chips for space applications; initiating weapon explosives with lasers; next-generation optoelectronics and microelectronics technology developments; chemometrics: new methods for improving chemical analysis; research team focuses ion beam to record-breaking intensities; standardizing the volt to quantum accuracy; new techniques improve robotic software development productivity; a practical laser plasma source for generating soft x-rays; exploring metal grain boundaries; massively parallel computing; modeling the amount of desiccant needed for moisture control; attacking pollution with sunshine; designing fuel-conversion catalysts with computers; extending a nuclear power plant's useful life; plasma-facing components for the International Thermonuclear Experimental Reactor.

Ion collector design for an energy recovery test proposal with the negative ion source NIO1

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

Variale, V., E-mail: vincenzo.variale@ba.infn.it; Cavenago, M.; Agostinetti, P.

2016-02-15

Commercial viability of thermonuclear fusion power plants depends also on minimizing the recirculation power used to operate the reactor. The neutral beam injector (NBI) remains one of the most important method for plasma heating and control. For the future fusion power plant project DEMO, a NBI wall plug efficiency at least of 0.45 is required, while efficiency of present NBI project is about 0.25. The D{sup −} beam from a negative ion source is partially neutralized by a gas cell, which leaves more than 40% of energy in residual beams (D{sup −} and D{sup +}), so that an ion beammore » energy recovery system can significantly contribute to optimize efficiency. Recently, the test negative ion source NIO1 (60 keV, 9 beamlets with 15 mA H{sup −} each) has been designed and built at RFX (Padua) for negative ion production efficiency and the beam quality optimization. In this paper, a study proposal to use the NIO1 source also for a beam energy recovery test experiment is presented and a preliminary design of a negative ion beam collector with simulations of beam energy recovery is discussed.« less

Development of a test rig for a helium twin-screw compressor

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

Wang, B. M.; Hu, Z. J.; Zhang, P.

2014-01-29

A large helium cryogenic system is being developed for use in great science projects, such as the International Thermonuclear Experimental Reactor (ITER), Large Helical Device (LHD), and the Experimental Advanced Superconducting Tokamak (EAST). In this cryogenic system, a twin-screw compressor is a key component. Therefore, it is necessary to obtain the compressor performance. To obtain the performance characteristics, a test rig for the compressor has been built. All the important performance parameters, including adiabatic efficiency, volumetric efficiency, oil injection characteristic, and noise characteristic can be acquired with the rig when sensors are installed in the test system. With the testmore » performance, the helium twin-screw compressor can be evaluated. Using these results, the design of the compressor can be improved.« less

Improving High-Temperature Measurements in Nuclear Reactors with Mo/Nb Thermocouples

NASA Astrophysics Data System (ADS)

Villard, J.-F.; Fourrez, S.; Fourmentel, D.; Legrand, A.

2008-10-01

Many irradiation experiments performed in research reactors are used to assess the effects of nuclear radiations on material or fuel sample properties, and are therefore a crucial stage in most qualification and innovation studies regarding nuclear technologies. However, monitoring these experiments requires accurate and reliable instrumentation. Among all measurement systems implemented in irradiation devices, temperature—and more particularly high-temperature (above 1000°C)—is a major parameter for future experiments related, for example, to the Generation IV International Forum (GIF) Program or the International Thermonuclear Experimental Reactor (ITER) Project. In this context, the French Commissariat à l’Energie Atomique (CEA) develops and qualifies innovative in-pile instrumentation for its irradiation experiments in current and future research reactors. Logically, a significant part of these research and development programs concerns the improvement of in-pile high-temperature measurements. This article describes the development and qualification of innovative high-temperature thermocouples specifically designed for in-pile applications. This key study has been achieved with technical contributions from the Thermocoax Company. This new kind of thermocouple is based on molybdenum and niobium thermoelements, which remain nearly unchanged by thermal neutron flux even under harsh nuclear environments, whereas typical high-temperature thermocouples such as Type C or Type S are altered by significant drifts caused by material transmutations under the same conditions. This improvement has a significant impact on the temperature measurement capabilities for future irradiation experiments. Details of the successive stages of this development are given, including the results of prototype qualification tests and the manufacturing process.

Lithium-based surfaces controlling fusion plasma behavior at the plasma-material interfacea)

NASA Astrophysics Data System (ADS)

Allain, Jean Paul; Taylor, Chase N.

2012-05-01

The plasma-material interface and its impact on the performance of magnetically confined thermonuclear fusion plasmas are considered to be one of the key scientific gaps in the realization of nuclear fusion power. At this interface, high particle and heat flux from the fusion plasma can limit the material's lifetime and reliability and therefore hinder operation of the fusion device. Lithium-based surfaces are now being used in major magnetic confinement fusion devices and have observed profound effects on plasma performance including enhanced confinement, suppression and control of edge localized modes (ELM), lower hydrogen recycling and impurity suppression. The critical spatial scale length of deuterium and helium particle interactions in lithium ranges between 5-100 nm depending on the incident particle energies at the edge and magnetic configuration. Lithium-based surfaces also range from liquid state to solid lithium coatings on a variety of substrates (e.g., graphite, stainless steel, refractory metal W/Mo/etc., or porous metal structures). Temperature-dependent effects from lithium-based surfaces as plasma facing components (PFC) include magnetohydrodynamic (MHD) instability issues related to liquid lithium, surface impurity, and deuterium retention issues, and anomalous physical sputtering increase at temperatures above lithium's melting point. The paper discusses the viability of lithium-based surfaces in future burning-plasma environments such as those found in ITER and DEMO-like fusion reactor devices.

FROM THE HISTORY OF PHYSICS: The extraordinarily beautiful physical principle of thermonuclear charge design (on the occasion of the 50th anniversary of the test of RDS-37 — the first Soviet two-stage thermonuclear charge)

NASA Astrophysics Data System (ADS)

Goncharov, German A.

2005-11-01

On 22 November 1955, the Semipalatinsk test site saw the test of the first domestic two-stage thermonuclear RDS-37 charge. The charge operation was based on the principle of radiation implosion. The kernel of the principle consists in the radiation generated in a primary A-bomb explosion and confined by the radiation-opaque casing propagating throughout the interior casing volume and flowing around the secondary thermonuclear unit. The secondary unit experiences a strong compression under the irradiation, with a resulting nuclear and thermonuclear explosion. The RDS-37 explosion was the strongest of all those ever realized at the Semipalatinsk test site. It produced an indelible impression on the participants in the test. This document-based paper describes the genesis of the ideas underlying the RDS-37 design and reflects the critical moments in its development. The advent of RDS-37 was an outstanding accomplishment of the scientists and engineers of our country.

Dealing With Russian Tactical Nuclear Weapons

DTIC Science & Technology

2004-01-01

the plants, facilities, and equipment necessary to safely secure and store weapons pits (the plutonium “triggers” at the center of a thermonuclear ... bomb , and hence the most critical piece) proffered by either party. Furthermore, the statute requires the IAEA to establish control over the pits until

Controlled Nuclear Fusion.

ERIC Educational Resources Information Center

Glasstone, Samuel

This publication is one of a series of information booklets for the general public published by The United States Atomic Energy Commission. Among the topics discussed are: Importance of Fusion Energy; Conditions for Nuclear Fusion; Thermonuclear Reactions in Plasmas; Plasma Confinement by Magnetic Fields; Experiments With Plasmas; High-Temperature…

Thermonuclear milestones: (2) Beginnings of the Soviet H-bomb program

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

Goncharov, G.A.

1996-11-01

Early Soviet theoretical work on thermonuclear ignition was adied by espionage, but many important ideas were conceived and developed independently {copyright} {ital 1996 American Institute of Physics.}

Effects of Welding Parameters on Mechanical Properties in Electron Beam Welded CuCrZr Alloy Plates

NASA Astrophysics Data System (ADS)

Jaypuria, Sanjib; Doshi, Nirav; Pratihar, Dilip Kumar

2018-03-01

CuCrZr alloys are attractive structural materials for plasma-facing components (PFC) and heat sink element in the International Thermonuclear Experimental Reactor (ITER) fusion reactors. This material has gained so much attention because of its high thermal conductivity and fracture toughness, high resistance to radiation damage and stability at elevated temperatures. The objective of this work is to study the effects of electron beam welding parameters on the mechanical strength of the butt welded CuCrZr joint. Taguchi method is used as the design of experiments to optimize the input parameters, such as accelerating voltage, beam current, welding speed, oscillation amplitude and frequency. The joint strength and ductility are the desired responses, which are measured through ultimate tensile strength and percent elongation, respectively. Accelerating voltage and welding speed are found to have significant influence on the strength. A combination of low amplitude and high-frequency oscillation is suggested for the higher joint strength and ductility. There is a close agreement between Taguchi predicted results and experimental ones. Fractographic analysis of joint and weld zone analysis are carried out to study the failure behaviour and microstructural variation in the weld zone, respectively.

Assessment and selection of materials for ITER in-vessel components

NASA Astrophysics Data System (ADS)

Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

2000-12-01

During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

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

Gorelenkov, Nikolai N

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

AC Josephson effect applications in microwave systems

NASA Astrophysics Data System (ADS)

Larkin, Serguey Y.

1996-12-01

A complication of the tasks solving by the modem radliolocation, radionavigation and communication systems connected with the demand promotion to the resolution and accuracy of coordinates definition and increase in the volumes of transmitted information in satellite communication systems has resulted in boisterous mastering of millimeter wave bands. Success in microwave technology reached in 80' allowed such leading instrument developing companies as Hewlett Packard; EIP, lB millimeter etc. to set up an output of mm- and submm-wave bands devices and systems. It has streamlined Scientific Technological Progress in several spheres, since millimeter, through infra-red frequency range was closed to researchers for a long period of time because of the absence of necessary equipment. At present microwave devices of the short-wave part of mm- wave band and of submm- wave bands are used not only in radiolocation and communications. Unique diagnostic systems based on the analysis of the radiation parameters of different microwave sources were created. They have their application in medicine, thermonuclear energetics, radioastronomy, biology, nuclear physics, the physics of the solid state body, geology, etc. The above circumstances caused the beginning of the measuring microwave technology researches in 60 to 600 GHz frequency range: generators, power and frequency meters, spectrum analyzers. The task of working out equipment and techniques of the effective control as well as frequency and intensity measurements of the microwave signals in the investigated range is of the special interest. Here are some examples. The creation of a thermonuclear reactor in ITER project is considered to be the project of the century in the energetics sphere. One of the basic engineering tasks in the course of project realization is the creation of the diagnostic equipment realizing in real time spectrum analysis of thermonuclear plasma radiation at the so called cyclotron hannonics. Such analysis allow to get the picture of temperature distribution along the plasma cord diameter in accordance with dynamics of thermonuclear process development. Modem raclioastronomic research gives scientists the unique information on the world tructure. It is also necessary to analyze Space microwave radiation providing exclusive sensitivity of the equipment. In both cases equipment is required to be superwide band, to have high sensitivity and ability to operate at more than 300 GHz frequencies. Today all these requirements are met by the devices using the ac Josephson effect. The Josephson junctions are used as an active transforming element in such devices. At the end of 20 century the sphere of their utilization embraces medicine, communications, radiophysics, space exploration, ecology, military use, etc. The State Research Center "Fonon" ( SRC "Fonon") of the State Committee on Science and Technology of Ukraine was founded in 1991. The main aim of its creation was to concentrate the scientific and financial efforts for development and production of unique devices based on the results of fundamental study in physics of high T superconductivity. First of all we were interested in technological research on the obtaining of low impedance Josephson junctions out of the High T thin films. Using such junctions in combination with our original techniques developed in our Center we have succeed in creating the following new generation equipment: industrial set-up of the frequency meter in the range of 60 ... 600 GHz; experimental set-up of the spectrum analyzer operating in the range of 50 250 GHz; experimental model of radiometric receiver in 180...260 GHz range. All the above devices are based on the using ac Josephson effect for the receiving and processing mm- and submm- microwave signals.

An Experimental Examination of the Loss-of-Flow Accident Phenomenon for Prototypical ITER Divertor Channels of Y = 0 and Y = 2

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

Marshall, Theron D.; McDonald, Jimmie M.; Cadwallader, Lee C.

2000-01-15

This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups' survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pumpmore » was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.« less

An experimental examination of the loss-of-flow accident phenomenon for prototypical ITER divertor channels of Y=0 and Y=2

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

Marshall, T.D.; McDonald, J.M.; Cadwallader, L.C.

2000-01-01

This paper discusses the thermal response of two prototypical International Thermonuclear Experimental Reactor (ITER) divertor channels during simulated loss-of-flow-accident (LOFA) experiments. The thermal response was characterized by the time-to-burnout (TBO), which is a figure of merit on the mockups' survivability. Data from the LOFA experiments illustrate that (a) the pre-LOFA inlet velocity does not significantly influence the TBO, (b) the incident heat flux (IHF) does influence the TBO, and (c) a swirl tape insert significantly improves the TBO and promotes the initiation of natural circulation. This natural circulation enabled the mockup to absorb steady-state IHFs after the coolant circulation pumpmore » was disabled. Several methodologies for thermal-hydraulic modeling of the LOFA were attempted.« less

Analysis of activation and shutdown contact dose rate for EAST neutral beam port

NASA Astrophysics Data System (ADS)

Chen, Yuqing; Wang, Ji; Zhong, Guoqiang; Li, Jun; Wang, Jinfang; Xie, Yahong; Wu, Bin; Hu, Chundong

2017-12-01

For the safe operation and maintenance of neutral beam injector (NBI), specific activity and shutdown contact dose rate of the sample material SS316 are estimated around the experimental advanced superconducting tokamak (EAST) neutral beam port. Firstly, the neutron emission intensity is calculated by TRANSP code while the neutral beam is co-injected to EAST. Secondly, the neutron activation and shutdown contact dose rates for the neutral beam sample materials SS316 are derived by the Monte Carlo code MCNP and the inventory code FISPACT-2007. The simulations indicate that the primary radioactive nuclides of SS316 are 58Co and 54Mn. The peak contact dose rate is 8.52 × 10-6 Sv/h after EAST shutdown one second. That is under the International Thermonuclear Experimental Reactor (ITER) design values 1 × 10-5 Sv/h.

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

Okabayashi, M.; Zanca, P.; Strait, E. J.

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

Nucleosynthesis in Thermonuclear Supernovae

NASA Astrophysics Data System (ADS)

Seitenzahl, Ivo Rolf; Townsley, Dean M.

The explosion energy of thermonuclear (type Ia) supernovae is derived from the difference in nuclear binding energy liberated in the explosive fusion of light "fuel" nuclei, predominantly carbon and oxygen, into more tightly bound nuclear "ash" dominated by iron and silicon group elements. The very same explosive thermonuclear fusion event is also one of the major processes contributing to the nucleosynthesis of the heavy elements, in particular the iron-group elements. For example, most of the iron and manganese in the sun and its planetary system were produced in thermonuclear supernovae. Here, we review the physics of explosive thermonuclear burning in carbon-oxygen white dwarf material and the methodologies utilized in calculating predicted nucleosynthesis from hydrodynamic explosion models. While the dominant explosion scenario remains unclear, many aspects of the nuclear combustion and nucleosynthesis are common to all models and must occur in some form in order to produce the observed yields. We summarize the predicted nucleosynthetic yields for existing explosion models, placing particular emphasis on characteristic differences in the nucleosynthetic signatures of the different suggested scenarios leading to type Ia supernovae. Following this, we discuss how these signatures compare with observations of several individual supernovae, remnants, and the composition of material in our galaxy and galaxy clusters.

Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

NASA Astrophysics Data System (ADS)

Shi, Xue-Ming; Peng, Xian-Jue

2016-09-01

Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR-E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR-E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR-E will be given.

The quest for a z-pinch based fusion energy source—a historical perspective

NASA Astrophysics Data System (ADS)

Sethian, John

1997-05-01

Ever since 1958, when Oscar Anderson observed copious neutrons emanating from a "magnetically self-constricted column of deuterium plasma," scientists have attempted to develop the simple linear pinch into a fusion power source. After all, simple calculations show that if one can pass a current of slightly less than 2 million amperes through a stable D-T plasma, then one could achieve not just thermonuclear break-even, but thermonuclear gain. Moreover, several reactor studies have shown that a simple linear pinch could be the basis for a very attractive fusion system. The problem is, of course, that the seemingly simple act of passing 2 MA through a stable pinch has proven to be quite difficult to accomplish. The pinch tends to disrupt due to instabilities, either by the m=0 (sausage) or m=1 (kink) modes. Curtailing the growth of these instabilities has been the primary thrust of z-pinch fusion research, and over the years a wide variety of formation techniques have been tried. The early pinches were driven by relatively slow capacitive discharges and were formed by imploding a plasma column. The advent of fast pulsed power technology brought on a whole new repertoire of formation techniques, including: fast implosions, laser or field-enhanced breakdown in a uniform volume of gas, a discharge inside a small capillary, a frozen deuterium fiber isolated by vacuum, and staged concepts in which one pinch implodes upon another. And although none of these have yet to be successful, some have come tantalizingly close. This paper will review the history of this four-decade long quest for fusion power.

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

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

NASA Astrophysics Data System (ADS)

Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

2017-11-01

The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification of numerical codes, all examined configurations of the MHD flow are also investigated numerically.

High pressure stability of lithium metatitanate and metazirconate: Insight from experiments & ab-initio calculations

NASA Astrophysics Data System (ADS)

Chitnis, Abhishek; Chakraborty, B.; Tripathi, B. M.; Tyagi, A. K.; Garg, Nandini

2018-02-01

Lithium metatitanate (LTO) and lithium metazirconate (LZO) are lithium rich ceramics which can be used as tritium breeder materials for thermonuclear reactors. In-situ x-ray diffraction and ab-initio studies at high pressure show that LTO has a higher bulk modulus than that of LZO. In fact these studies indicate that they are the least compressible of the known lithium rich ceramics like Li2O or Li4SiO4, which are potential candidates for blanket materials. These studies show that the TiO6 octahedra are responsible for the higher bulk modulus of LTO when compared to that of LZO. It has also been shown that the compressibility and distortion of the softer LiO6 octahedra can be controlled by altering the stacking sequence of the more rigid covalently bonded octahedra. This knowledge can be used by chemists to design new lithium based ceramics with higher bulk modulus. It was observed that LTO was stable upto 34 GPa. Ab initio DFT calculations helped to understand the anisotropy in compressibility of both LZO and LTO. This study also shows, that even though the empirical potentials developed by Vijaykumar et al. successfully determine the ambient pressure structure of lithium metatitanate, they cannot be used at non ambient conditions like high pressure [1].

Nano-size metallic oxide particle synthesis in Fe-Cr alloys by ion implantation

NASA Astrophysics Data System (ADS)

Zheng, C.; Gentils, A.; Ribis, J.; Borodin, V. A.; Delauche, L.; Arnal, B.

2017-10-01

Oxide Dispersion Strengthened (ODS) steels reinforced with metal oxide nanoparticles are advanced structural materials for nuclear and thermonuclear reactors. The understanding of the mechanisms involved in the precipitation of nano-oxides can help in improving mechanical properties of ODS steels, with a strong impact for their commercialization. A perfect tool to study these mechanisms is ion implantation, where various precipitate synthesis parameters are under control. In the framework of this approach, high-purity Fe-10Cr alloy samples were consecutively implanted with Al and O ions at room temperature and demonstrated a number of unexpected features. For example, oxide particles of a few nm in diameter could be identified in the samples already after ion implantation at room temperature. This is very unusual for ion beam synthesis, which commonly requires post-implantation high-temperature annealing to launch precipitation. The observed particles were composed of aluminium and oxygen, but additionally contained one of the matrix elements (chromium). The crystal structure of aluminium oxide compound corresponds to non-equilibrium cubic γ-Al2O3 phase rather than to more common corundum. The obtained experimental results together with the existing literature data give insight into the physical mechanisms involved in the precipitation of nano-oxides in ODS alloys.

Conductor analysis of the ITER FEAT poloidal field coils during a plasma scenario

NASA Astrophysics Data System (ADS)

Nicollet, S.; Hertout, P.; Duchateau, J. L.; Bleyer, A.; Bessette, D.

2002-05-01

In the framework of the ITER (International Thermonuclear Experimental Reactor) FEAT (Fusion Energy Advanced Tokamak) project, a fully superconducting PF (Poloidal Field) system has been designed in detail. The Central Solenoid and the 6 equilibrium coils constituting the PF system provide the magnetic fields which develop, shape and control the 15 MA plasma during the 1800 s of a typical plasma scenario. The 6 PF coils will be wound two-in-hand from a 45 kA niobium-titanium CICC (Cable-In-Conduit-Conductor). These coils will experience severe heat loads specially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation and AC losses (30 to 300 kJ). The AC losses along the PF coil pancakes are deduced from accurate magnetic field computations performed with a 3D magnetostatic code, TRAPS. The nuclear heating and the thermal radiation are assumed to be uniform over a given face of the PF coils. These heat loads are used as input to perform the thermal and hydraulic analysis with a finite element code, GANDALF. The temperature increases (0.1 to 0.4 K) are computed, the margins and performances of the conductor are evaluated.

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

Barnett, C.F.; Gauster, W.B.; Ray, J.A.

A graphical compilation is presented of atomic and molecular cross sections of interest to controlled thermonuclear research. The cross sections are shown, as a function of energy, for collision processes involving molecular ion dissociation, charge exchange, excitation, ionization, photoionization, scattering, energy loss, and recombination. Pertinent nuclear cross sections are also included. A bibliography is given covering the literature since 1950. (auth)

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

Hanada, M., E-mail: hanada.masaya@jaea.go.jp; Kojima, A.; Tobari, H.

In order to realize negative ion sources and accelerators to be applicable to International Thermonuclear Experimental Reactor and JT-60 Super Advanced, a large cesium (Cs)-seeded negative ion source and a multi-aperture and multi-stage electric acceleration have been developed at Japan Atomic Energy Agency (JAEA). Long pulse production and acceleration of the negative ion beams have been independently carried out. The long pulse production of the high current beams has achieved 100 s at the beam current of 15 A by modifying the JT-60 negative ion source. The pulse duration time is increased three times longer than that before the modification.more » As for the acceleration, a pulse duration time has been also extended two orders of magnitudes from 0.4 s to 60 s. The developments of the negative ion source and acceleration at JAEA are well in progress towards the realization of the negative ion sources and accelerators for fusion applications.« less

On the safety of ITER accelerators.

PubMed

Li, Ge

2013-01-01

Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate -1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER.

Carbon fiber composites application in ITER plasma facing components

NASA Astrophysics Data System (ADS)

Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

1998-10-01

Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

Characterization of 316L(N)-IG SS joint produced by hot isostatic pressing technique

NASA Astrophysics Data System (ADS)

Nakano, J.; Miwa, Y.; Tsukada, T.; Kikuchi, M.; Kita, S.; Nemoto, Y.; Tsuji, H.; Jitsukawa, S.

2002-12-01

Type 316L(N) stainless steel of the international thermonuclear experimental reactor grade (316L(N)-IG SS) is being considered for the first wall/blanket module. Hot isostatic pressing (HIP) technique is expected for the fabrication of the module. To evaluate the integrity and susceptibility to stress corrosion cracking (SCC) of HIPed 316L(N)-IG SS, tensile tests in vacuum and slow strain rate tests in high temperature water were performed. Specimen with the HIPed joint had similar tensile properties to specimens of 316L(N)-IG SS, and did not show susceptibility to SCC in oxygenated water at 423 K. Thermally sensitized specimen was low susceptible to SCC even in the creviced condition. It is concluded that the tensile properties of HIPed SS are as high as those of the base alloy and the HIP process caused no deleterious effects.

On the safety of ITER accelerators

PubMed Central

Li, Ge

2013-01-01

Three 1 MV/40A accelerators in heating neutral beams (HNB) are on track to be implemented in the International Thermonuclear Experimental Reactor (ITER). ITER may produce 500 MWt of power by 2026 and may serve as a green energy roadmap for the world. They will generate −1 MV 1 h long-pulse ion beams to be neutralised for plasma heating. Due to frequently occurring vacuum sparking in the accelerators, the snubbers are used to limit the fault arc current to improve ITER safety. However, recent analyses of its reference design have raised concerns. General nonlinear transformer theory is developed for the snubber to unify the former snubbers' different design models with a clear mechanism. Satisfactory agreement between theory and tests indicates that scaling up to a 1 MV voltage may be possible. These results confirm the nonlinear process behind transformer theory and map out a reliable snubber design for a safer ITER. PMID:24008267

Technology of Producing the Contact Connections of Superconductor Metal-Sheathed Cable

NASA Astrophysics Data System (ADS)

Jakubowski, Andrzej

2017-06-01

The technology of producing the current contact connections on the superconductor cable edges is presented. This lead cable is used as one of the major elements of the magnetic system in thermonuclear reactor construction, actuality for modern world energy. The technology is realized by the radial draft of metal thin-walled tube on the conductor's package. The filling of various profiles by round section wire is optimized. Geometrical characteristics of the dangerous crosssection (as a broken ring) of thin-walled tube injured by the sector cut-out are accounted. The comparative strength calculation of the solid and injured tubes at a longitudinal compression and lateral bending is acted. The radial draft mechanism of cylindrical thin-walled sheath with the wire packing is designed. The necessity to use the nonlinear theory for the sheaths calculate is set. The resilient co-operation of wires as the parallel located cylinders with the contact stripes of rectangular form is considered.

Influence of the optical fiber type on the performances of fiber-optics current sensor dedicated to plasma current measurement in ITER.

PubMed

Aerssens, Matthieu; Descamps, Frédéric; Gusarov, Andrei; Mégret, Patrice; Moreau, Philippe; Wuilpart, Marc

2015-07-01

In this paper, we compare, by means of simulations using the Jones formalism, the performances of several optical fiber types (low birefringence and spun fibers) for the measurement of plasma current in international thermonuclear experimental reactor (ITER). The main results presented in this paper concern the minimum value of the ratio between the beat length and the spun period, which allows meeting the ITER current measurement specifications. Assuming a high-birefringence spun fiber with a beat length of 3 mm, we demonstrate that the minimum ratio between the beat length and the spun period is 4.4 when considering a 28 m long sensing fiber surrounding the vacuum vessel. This minimum ratio rises to 10.14 when a 100 m long lead fiber connecting the interrogating system to the sensing fiber is taken into account.

Recent Updates to the MELCOR 1.8.2 Code for ITER Applications

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

Merrill, Brad J

This report documents recent changes made to the MELCOR 1.8.2 computer code for application to the International Thermonuclear Experimental Reactor (ITER), as required by ITER Task Agreement ITA 81-18. There are four areas of change documented by this report. The first area is the addition to this code of a model for transporting HTO. The second area is the updating of the material oxidation correlations to match those specified in the ITER Safety Analysis Data List (SADL). The third area replaces a modification to an aerosol tranpsort subroutine that specified the nominal aerosol density internally with one that now allowsmore » the user to specify this density through user input. The fourth area corrected an error that existed in an air condensation subroutine of previous versions of this modified MELCOR code. The appendices of this report contain FORTRAN listings of the coding for these modifications.« less

Dirac R -matrix calculations for the electron-impact excitation of neutral tungsten providing noninvasive diagnostics for magnetic confinement fusion

NASA Astrophysics Data System (ADS)

Smyth, R. T.; Ballance, C. P.; Ramsbottom, C. A.; Johnson, C. A.; Ennis, D. A.; Loch, S. D.

2018-05-01

Neutral tungsten is the primary candidate as a wall material in the divertor region of the International Thermonuclear Experimental Reactor (ITER). The efficient operation of ITER depends heavily on precise atomic physics calculations for the determination of reliable erosion diagnostics, helping to characterize the influx of tungsten impurities into the core plasma. The following paper presents detailed calculations of the atomic structure of neutral tungsten using the multiconfigurational Dirac-Fock method, drawing comparisons with experimental measurements where available, and includes a critical assessment of existing atomic structure data. We investigate the electron-impact excitation of neutral tungsten using the Dirac R -matrix method, and by employing collisional-radiative models, we benchmark our results with recent Compact Toroidal Hybrid measurements. The resulting comparisons highlight alternative diagnostic lines to the widely used 400.88-nm line.

Investigation of the effect of Alfven resonance absorption on fast wave current drive in ITER

NASA Astrophysics Data System (ADS)

Alava, M. J.; Heikkinen, J. A.; Hellsten, T.

The use of frequencies below the ion cyclotron frequency of minority ion species or second harmonic of majority species has been proposed for fast wave current drive in order to reduce or to avoid ion cyclotron damping. For these scenarios, the Alfven resonance can appear on the high field side of a tokamak. The presence of this resonance causes parasitic absorption competing with the electron Landau damping and transit time magnetic pumping responsible for the fast wave current drive. In the present study, the mode conversion at the Alfven resonance is shown to be of the order of 5 to 10 percent in the current drive scenarios for the planned International Thermonuclear Experimental Reactor (ITER) experiment. However, if the single pass absorption in the center can be made sufficiently high, the conversion at the Alfven resonance becomes negligible.

Tritium saturation in plasma-facing materials surfaces1

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Anderl, Robert A.; Causey, Rion A.; Federici, Gianfranco; Haasz, Anthony A.; Pawelko, Robert J.

1998-10-01

Plasma-facing components in the International Thermonuclear Experimental Reactor (ITER) will experience high heat loads and intense plasma fluxes of order 10 20-10 23 particles/m 2s. Experiments on Be and W, two of the materials considered for use in ITER, have revealed that a tritium saturation phenomenon can take place under these conditions in which damage to the surface results that enhances the return of implanted tritium to the plasma and inhibits uptake of tritium. This phenomenon is important because it implies that tritium inventories due to implantation in these plasma-facing materials will probably be lower than was previously estimated using classical recombination-limited release at the plasma surface. Similarly, permeation through these components to the coolant streams should be reduced. In this paper we discuss evidences for the existence of this phenomenon, describe techniques for modeling it, and present results of the application of such modeling to prior experiments.

The physics of tokamak start-up

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

Mueller, D.

Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in themore » solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.« less

Numerical verification of bounce-harmonic resonances in neoclassical toroidal viscosity for tokamaks.

PubMed

Kim, Kimin; Park, Jong-Kyu; Boozer, Allen H

2013-05-03

This Letter presents the first numerical verification for the bounce-harmonic (BH) resonance phenomena of the neoclassical transport in a tokamak perturbed by nonaxisymmetric magnetic fields. The BH resonances were predicted by analytic theories of neoclassical toroidal viscosity (NTV), as the parallel and perpendicular drift motions can be resonant and result in a great enhancement of the radial momentum transport. A new drift-kinetic δf guiding-center particle code, POCA, clearly verified that the perpendicular drift motions can reduce the transport by phase-mixing, but in the BH resonances the motions can form closed orbits and particles radially drift out fast. The POCA calculations on resulting NTV torque are largely consistent with analytic calculations, and show that the BH resonances can easily dominate the NTV torque when a plasma rotates in the perturbed tokamak and therefore, is a critical physics for predicting the rotation and stability in the International Thermonuclear Experimental Reactor.

Assessment of possible failure modes and non-destructive examination of the ITER pre-compression rings

NASA Astrophysics Data System (ADS)

Knaster, J.; Evans, D.; Rajainmaki, H.

2012-06-01

The pre-compression rings (PCRs) for the International Thermonuclear Experimental Reactor (ITER) represent one of the largest and most highly stressed composite structures ever designed for long term operation at 4K. Three rings, each 5m in diameter and 337 × 288 mm in cross-section, will be installed at the top and bottom of the eighteen "D" shaped Toroidal Field (TF) coils to apply a total centripetal load of 70 MN per TF coil. The interaction of the 68 kA conductor current circulating in the coil (for a total of 9.1MA) with the required magnetic field to confine the plasma during operation will result in Lorentz forces that build in-plane and out-of-plane loads. The PCRs are essential to keep the stresses below the acceptable level for the ITER magnets structural materials.

Investigating the Neutral-Gas Manometers in the Wendelstein 7-X Experimental Fusion Reactor

NASA Astrophysics Data System (ADS)

Maisano-Brown, Jeannette; Wenzel, Uwe; Sunn-Pederson, Thomas

2017-01-01

The neutral-gas manometer is a powerful diagnostic tool used in the Wendelstein 7-X stellarator, a magnetized fusion experiment located in Germany. The Wendelstein, produced at a cost of 1.2 billion euros, and 20 years in the making, had its first experimental results in Winter 2016. Initial findings exceeded expectations but further study is still necessary. The particular instrument we examined was a hot-cathode ionization gauge, critical for attaining a quality in-vessel environment and a stable plasma. However, after the winter operation of Wendelstein, we found that some of the gauges had failed the six-second (maximum) plasma runs. Wendelstein is on track for 30-minute operations within three years, so it has become of utmost importance to scrutinize gauge design claims. We therefore subjected the devices to high magnetic field, input current, and temperature, as well as to long operational periods. Our results confirmed that the manometer cannot survive a 30-minute run. Though our findings did motivate promising recommendations for design improvement and for further experimentation so that the gauge can be ready for upcoming operations in Summer 2017 and eventual installment in ITER, the International Thermonuclear Experimental Reactor, currently under construction. This research was graciously supported by the Max Planck Institute and the MIT-Germany Initiative.

Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

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

Linke, J.; Bolt. H.; Breitbach, G.

1994-12-31

To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors such as ITER, neutron irradiation and subsequent high heat flux tests will be most essential. The effect of neutron damage will be simulated in material test reactors (such as the HFR-Petten) in a fission neutron environment. To investigate the heat loads during normal and off-normal operation scenarios a 60 kW electron beam test stand (Juelich Divertor Test Facility in Hot Cells, JUDITH) has been installed in a hot cell which can be operated by remote handling techniques. In this facilitymore » inertially cooled test coupons can be handled as well as small actively cooled divertor mock-ups. A special clamping mechanism for small test coupons (25 mm x 25 mm x 35 mm) with an integrated coolant channel within a copper or TZM heat sink has been developed and tested in an electron beam test bed. This method is an attractive alternative to costly large scale tests on complete divertor modules. The temperature and stress fields in individual CFC or beryllium tiles brazed to metallic heat sink (e.g. copper or TZM) can be investigated before and after neutron irradiation with moderate efforts.« less

Merging white dwarfs and thermonuclear supernovae.

PubMed

van Kerkwijk, M H

2013-06-13

Thermonuclear supernovae result when interaction with a companion reignites nuclear fusion in a carbon-oxygen white dwarf, causing a thermonuclear runaway, a catastrophic gain in pressure and the disintegration of the whole white dwarf. It is usually thought that fusion is reignited in near-pycnonuclear conditions when the white dwarf approaches the Chandrasekhar mass. I briefly describe two long-standing problems faced by this scenario, and the suggestion that these supernovae instead result from mergers of carbon-oxygen white dwarfs, including those that produce sub-Chandrasekhar-mass remnants. I then turn to possible observational tests, in particular, those that test the absence or presence of electron captures during the burning.

Two-dimensional over-all neutronics analysis of the ITER device

NASA Astrophysics Data System (ADS)

Zimin, S.; Takatsu, Hideyuki; Mori, Seiji; Seki, Yasushi; Satoh, Satoshi; Tada, Eisuke; Maki, Koichi

1993-07-01

The present work attempts to carry out a comprehensive neutronics analysis of the International Thermonuclear Experimental Reactor (ITER) developed during the Conceptual Design Activities (CDA). The two-dimensional cylindrical over-all calculational models of ITER CDA device including the first wall, blanket, shield, vacuum vessel, magnets, cryostat and support structures were developed for this purpose with a help of the DOGII code. Two dimensional DOT 3.5 code with the FUSION-40 nuclear data library was employed for transport calculations of neutron and gamma ray fluxes, tritium breeding ratio (TBR), and nuclear heating in reactor components. The induced activity calculational code CINAC was employed for the calculations of exposure dose rate after reactor shutdown around the ITER CDA device. The two-dimensional over-all calculational model includes the design specifics such as the pebble bed Li2O/Be layered blanket, the thin double wall vacuum vessel, the concrete cryostat integrated with the over-all ITER design, the top maintenance shield plug, the additional ring biological shield placed under the top cryostat lid around the above-mentioned top maintenance shield plug etc. All the above-mentioned design specifics were included in the employed calculational models. Some alternative design options, such as the water-rich shielding blanket instead of lithium-bearing one, the additional biological shield plug at the top zone between the poloidal field (PF) coil No. 5, and the maintenance shield plug, were calculated as well. Much efforts have been focused on analyses of obtained results. These analyses aimed to obtain necessary recommendations on improving the ITER CDA design.

Thermonuclear Fusion: An Energy Source for the Future

ERIC Educational Resources Information Center

Drummond, William E.

1973-01-01

Discusses current research in thermonuclear fusion with particular emphasis on the problem of confining hot plasma. Recent experiments indicate that magnetic bottles called tokamaks may achieve the necessary confinement times, and this break-through has given renewed optimism to the feasibility of commercial fusion power by the turn of the…

Gamma-Ray Effects Testing in Lawrence Livermore National Laboratory’s Nova Upgrade Facility

DTIC Science & Technology

1992-03-01

However, the distribution of energy between neutrons and photons from the ICF pellet will be different then that from a thermonuclear bomb . In a...of the neutron energy to photons. Photons make up most of the energy released from a thermonuclear bomb (Glasstone, S., 1977:340-342). The neutrons

Rocket propulsion by thermonuclear micro-bombs ignited with intense relativistic electron beams.

NASA Technical Reports Server (NTRS)

Winterberg, F.

1971-01-01

Discussion of a method for the ignition of a thermonuclear microbomb by means of an intense relativistic electron beam with regard to its potential application to rocket propulsion. With such a system, exhaust velocities up to 1000 km/sec, corresponding to a specific impulse of 100,000 sec, seem to be within the realm of possibility. The rocket is propelled by a chain of thermonuclear microbombs exploded in a concave magnetic mirror produced by superconducting field coils. The magnetic pressure of the field reflects the fireball generated by the explosion. For the large capacitor bank required to generate the intense relativistic electron beam, a desirable lightweight design may be possible through use of ferroelectric materials. Because of the high cost of the T-D and He 3-D thermonuclear material, the system has to be optimized by minimizing the T-D and He 3-D consumption by a proper TD and He 3-D fuel to hydrogen propellant mass ratio, leading to a larger total system mass than would be absolutely necessary.

FAILURE OF A NEUTRINO-DRIVEN EXPLOSION AFTER CORE-COLLAPSE MAY LEAD TO A THERMONUCLEAR SUPERNOVA

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

Kushnir, Doron; Katz, Boaz, E-mail: kushnir@ias.edu

We demonstrate that ∼10 s after the core-collapse of a massive star, a thermonuclear explosion of the outer shells is possible for some (tuned) initial density and composition profiles, assuming that the neutrinos failed to explode the star. The explosion may lead to a successful supernova, as first suggested by Burbidge et al. We perform a series of one-dimensional (1D) calculations of collapsing massive stars with simplified initial density profiles (similar to the results of stellar evolution calculations) and various compositions (not similar to 1D stellar evolution calculations). We assume that the neutrinos escaped with a negligible effect on themore » outer layers, which inevitably collapse. As the shells collapse, they compress and heat up adiabatically, enhancing the rate of thermonuclear burning. In some cases, where significant shells of mixed helium and oxygen are present with pre-collapsed burning times of ≲100 s (≈10 times the free-fall time), a thermonuclear detonation wave is ignited, which unbinds the outer layers of the star, leading to a supernova. The energy released is small, ≲10{sup 50} erg, and negligible amounts of synthesized material (including {sup 56}Ni) are ejected, implying that these 1D simulations are unlikely to represent typical core-collapse supernovae. However, they do serve as a proof of concept that the core-collapse-induced thermonuclear explosions are possible, and more realistic two-dimensional and three-dimensional simulations are within current computational capabilities.« less

Theoretical Issues for Plasma Regimes to be Explored by the Ignitor Experiment

NASA Astrophysics Data System (ADS)

Cardinali, A.; Coppi, B.; Sonnino, G.

2014-10-01

At present, the Ignitor experiment is the only one designed and planned to approach and explore ignition regimes under controlled DT burning conditions. The machine parameters have been established on the basis of existing knowledge of the confinement properties of high density plasmas. A variety of improved confinement regimes are expected to be accessible by means of the available ICRH heating power in addition to the prevalent programmable Ohmic heating power and relying on the injection of high velocity pellets for density profile control. The relevance of the various known confinement regimes to the objectives of Ignitor is discussed. Among other theoretical efforts, a non-linear thermal energy balance equation is investigated to study the onset of thermonuclear instability in the plasmas expected to be produced in Ignitor. The equation for the temperature profile in the equilibrium state is solved with the resulting profiles in agreement with those obtained by a full transport code and commonly adopted scalings for them. The evolution of the thermonuclear instability that relies on the solution of the time dependent energy balance equation is obtained. Sponsored in part by the U.S. DOE.

Fusion plasma theory project summaries

NASA Astrophysics Data System (ADS)

1993-10-01

This Project Summary book is a published compilation consisting of short descriptions of each project supported by the Fusion Plasma Theory and Computing Group of the Advanced Physics and Technology Division of the Department of Energy, Office of Fusion Energy. The summaries contained in this volume were written by the individual contractors with minimal editing by the Office of Fusion Energy. Previous summaries were published in February of 1982 and December of 1987. The Plasma Theory program is responsible for the development of concepts and models that describe and predict the behavior of a magnetically confined plasma. Emphasis is given to the modelling and understanding of the processes controlling transport of energy and particles in a toroidal plasma and supporting the design of the International Thermonuclear Experimental Reactor (ITER). A tokamak transport initiative was begun in 1989 to improve understanding of how energy and particles are lost from the plasma by mechanisms that transport them across field lines. The Plasma Theory program has actively participated in this initiative. Recently, increased attention has been given to issues of importance to the proposed Tokamak Physics Experiment (TPX). Particular attention has been paid to containment and thermalization of fast alpha particles produced in a burning fusion plasma as well as control of sawteeth, current drive, impurity control, and design of improved auxiliary heating. In addition, general models of plasma behavior are developed from physics features common to different confinement geometries. This work uses both analytical and numerical techniques. The Fusion Theory program supports research projects at U.S. government laboratories, universities and industrial contractors. Its support of theoretical work at universities contributes to the office of Fusion Energy mission of training scientific manpower for the U.S. Fusion Energy Program.

Terrorist Use of the Internet and Related Information Technologies

DTIC Science & Technology

2002-05-15

the Anarchist, Vol I," 4 January 1985 [online]; available from <http://www.etext.org/CuD/Misc/anarch>. 33 See _____, "How to Build a Thermonuclear Bomb ," About...Development, no date. [Online]; previously available from <http://www.hlf.org> (accessed 6 December 2001). _____. "How to Build a Thermonuclear Bomb ." About

Particle and momentum confinement in tokamak plasmas with unbalanced neutral beam injection and strong rotation

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

Malik, M.A.

1988-01-01

There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such asmore » INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism.« less

Inertial Confinement Fusion as an Extreme Example of Dynamic Compression

NASA Astrophysics Data System (ADS)

Moses, E.

2013-06-01

Initiating and controlling thermonuclear burn at the national ignition facility (NIF) will require the manipulation of matter to extreme energy densities. We will discuss recent advances in both controlling the dynamic compression of ignition targets and our understanding of the physical states and processes leading to ignition. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory in part under Contract W-7405-Eng-48 and in part under Contract DE-AC52-07NA27344.

Basic requirements for a 1000-MW(electric) class tokamak fusion-fission hybrid reactor and its blanket concept

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

Hatayama, Ariyoshi; Ogasawara, Masatada; Yamauchi, Michinori

1994-08-01

Plasma size and other basic performance parameters for 1000-MW(electric) power production are calculated with the blanket energy multiplication factor, the M value, as a parameter. The calculational model is base don the International Thermonuclear Experimental Reactor (ITER) physics design guidelines and includes overall plant power flow. Plasma size decreases as the M value increases. However, the improvement in the plasma compactness and other basic performance parameters, such as the total plant power efficiency, becomes saturated above the M = 5 to 7 range. THus, a value in the M = 5 to 7 range is a reasonable choice for 1000-MW(electric)more » hybrids. Typical plasma parameters for 1000-MW(electric) hybrids with a value of M = 7 are a major radius of R = 5.2 m, minor radius of a = 1.7 m, plasma current of I{sub p} = 15 MA, and toroidal field on the axis of B{sub o} = 5 T. The concept of a thermal fission blanket that uses light water as a coolant is selected as an attractive candidate for electricity-producing hybrids. An optimization study is carried out for this blanket concept. The result shows that a compact, simple structure with a uniform fuel composition for the fissile region is sufficient to obtain optimal conditions for suppressing the thermal power increase caused by fuel burnup. The maximum increase in the thermal power is +3.2%. The M value estimated from the neutronics calculations is {approximately}7.0, which is confirmed to be compatible with the plasma requirement. These studies show that it is possible to use a tokamak fusion core with design requirements similar to those of ITER for a 1000-MW(electric) power reactor that uses existing thermal reactor technology for the blanket. 30 refs., 22 figs., 4 tabs.« less

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

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

NONE

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

Characterization of high flux magnetized helium plasma in SCU-PSI linear device

NASA Astrophysics Data System (ADS)

Xiaochun, MA; Xiaogang, CAO; Lei, HAN; Zhiyan, ZHANG; Jianjun, WEI; Fujun, GOU

2018-02-01

A high-flux linear plasma device in Sichuan University plasma-surface interaction (SCU-PSI) based on a cascaded arc source has been established to simulate the interactions between helium and hydrogen plasma with the plasma-facing components in fusion reactors. In this paper, the helium plasma has been characterized by a double-pin Langmuir probe. The results show that the stable helium plasma beam with a diameter of 26 mm was constrained very well at a magnetic field strength of 0.3 T. The core density and ion flux of helium plasma have a strong dependence on the applied current, magnetic field strength and gas flow rate. It could reach an electron density of 1.2 × 1019 m-3 and helium ion flux of 3.2 × 1022 m-2 s-1, with a gas flow rate of 4 standard liter per minute, magnetic field strength of 0.2 T and input power of 11 kW. With the addition of -80 V applied to the target to increase the helium ion energy and the exposure time of 2 h, the flat top temperature reached about 530 °C. The different sizes of nanostructured fuzz on irradiated tungsten and molybdenum samples surfaces under the bombardment of helium ions were observed by scanning electron microscopy. These results measured in the SCU-PSI linear device provide a reference for International Thermonuclear Experimental Reactor related PSI research.

Superconductivity and fusion energy—the inseparable companions

NASA Astrophysics Data System (ADS)

Bruzzone, Pierluigi

2015-02-01

Although superconductivity will never produce energy by itself, it plays an important role in energy-related applications both because of its saving potential (e.g., power transmission lines and generators), and its role as an enabling technology (e.g., for nuclear fusion energy). The superconducting magnet’s need for plasma confinement has been recognized since the early development of fusion devices. As long as the research and development of plasma burning was carried out on pulsed devices, the technology of superconducting fusion magnets was aimed at demonstrations of feasibility. In the latest generation of plasma devices, which are larger and have longer confinement times, the superconducting coils are a key enabling technology. The cost of a superconducting magnet system is a major portion of the overall cost of a fusion plant and deserves significant attention in the long-term planning of electricity supply; only cheap superconducting magnets will help fusion get to the energy market. In this paper, the technology challenges and design approaches for fusion magnets are briefly reviewed for past, present, and future projects, from the early superconducting tokamaks in the 1970s, to the current ITER (International Thermonuclear Experimental Reactor) and W7-X projects and future DEMO (Demonstration Reactor) projects. The associated cryogenic technology is also reviewed: 4.2 K helium baths, superfluid baths, forced-flow supercritical helium, and helium-free designs. Open issues and risk mitigation are discussed in terms of reliability, technology, and cost.

Fusion energy

NASA Astrophysics Data System (ADS)

1990-09-01

The main purpose of the International Thermonuclear Experimental Reactor (ITER) is to develop an experimental fusion reactor through the united efforts of many technologically advanced countries. The ITER terms of reference, issued jointly by the European Community, Japan, the USSR, and the United States, call for an integrated international design activity and constitute the basis of current activities. Joint work on ITER is carried out under the auspices of the International Atomic Energy Agency (IAEA), according to the terms of quadripartite agreement reached between the European Community, Japan, the USSR, and the United States. The site for joint technical work sessions is at the Max Planck Institute of Plasma Physics. Garching, Federal Republic of Germany. The ITER activities have two phases: a definition phase performed in 1988 and the present design phase (1989 to 1990). During the definition phase, a set of ITER technical characteristics and supporting research and development (R and D) activities were developed and reported. The present conceptual design phase of ITER lasts until the end of 1990. The objectives of this phase are to develop the design of ITER, perform a safety and environmental analysis, develop site requirements, define future R and D needs, and estimate cost, manpower, and schedule for construction and operation. A final report will be submitted at the end of 1990. This paper summarizes progress in the ITER program during the 1989 design phase.

Chemically ignited thermonuclear reactions—A near-term means for a high specific impulse—High thrust propulsion system

NASA Astrophysics Data System (ADS)

Winterberg, F.

The combination of metallic shells imploded with chemical explosives and the recently proposed magnetic booster target inertial fusion concept, could make possible the fissionless ignition of small thermonuclear explosions. In the magnetic booster concept a very dense but magnetically confined thermonuclear plasma of low yield serves as the trigger for an inertially confined thermonuclear plasma of high yield. For the most easily ignitable fusion reaction, the DT reaction, this could lead to a fissionless bomb propulsion system, with the advantage to have a much smaller yield of the pure fusion bombs as compared to either fission- or fission-induced fusion bombs, previously proposed for propulsion. Typically, the proposed propulsion concept should give a specific impulse of ˜ 3000 secs, corresponding to an exhaust velocity of ˜ 30 km/sec. If the energy released in each pure fusion bomb is of the order of 10 18 erg or the order of 100 tons of TNT, and if one fusion explosion per second takes place, the average thrust is of the order 10 3 tons. The propulsion system appears ideally suited for the fast economical transport of large spacecraft within the solar system.

Experiments with lithium limiter on T-11M tokamak and applications of the lithium capillary-pore system in future fusion reactor devices

NASA Astrophysics Data System (ADS)

Mirnov, S. V.; Azizov, E. A.; Evtikhin, V. A.; Lazarev, V. B.; Lyublinski, I. E.; Vertkov, A. V.; Prokhorov, D. Yu

2006-06-01

The paper is an overview of recent results of Li limiter testing in T-11M tokamak. The lithium limiter is based on the capillary-pore system (CPS) concept. The Li erosion process and deuterium (D2) and helium (He) sorption by Li first wall were investigated. The ability of capillary forces to confine the liquid Li in the CPS limiter during disruption was demonstrated. The idea of combined lithium limiter with thin (0.6 mm) CPS coating as a solution of the heat removal problem was realized. As a result the quasi steady-state tokamak regime with duration up to 0.3 s and clean (Zeff = 1) deuterium plasma has been achieved. The temporal evolution of the lithium surface temperature during discharge was measured by a IR radiometer and then was recalculated to the surface power load. For the estimation of the Li limiter erosion the Li neutral and ions spectral line emission were observed. The increase in lithium erosion as a result of limiter heating was discovered. The radial distribution of plasma column radiation measurements showed up to 90% of the total radiation losses in a relatively thin (5 cm) boundary layer and only 10% in a plasma centre during discharges with high Li influx. Oscillations of Li emission and saw-tooth-like oscillations of the limiter surface temperature have been detected in discharge regimes with highest Li limiter temperature (>600 °C). A version of Li CPS first wall of DEMO reactor and Li CPS limiter experiment in the International Thermonuclear Energy Reactor are suggested.

Electron Impact Cross Sections for Molecular Lasers

DTIC Science & Technology

1984-04-27

range coumunication and surveillance, isotope separation, and controlled thermonuclear fussion . Among all kinds of lasers, the gaseous discharge...shape resonance of n symmetry (reviewed by Schulz, 1976). Like vibrational excitation from the ground state, such process from nuclear -excited states as...energy range specifically in the 1-4 eV resonant region. 4 - A. Vibrational Excitation of Nuclear -Excited N2 For vibrational excitation by

On Coulomb collisions in the solar wind

NASA Astrophysics Data System (ADS)

Hellinger, P.; Travnicek, P. M.

2009-04-01

Collisional transport in anisotropic plasmas is investigated comparing theoretical predictions of the Fokker-Planck equation for bi-Maxwellian particle distribution functions (Kogan, 1961; Lehner, 1967) and results of the corresponding Langevin equation. References: Kogan, V. I., in Plasma Physics and the Problem of Controlled Thermonuclear Reactions, edited by M. A. Leontovich, Pergamon Press, New York, , vol. 1, 153, 1961. Lehner, G., Zeitschrift fur Physik, 206, 284, 1967.

Demonstration of thermonuclear conditions in magnetized liner inertial fusion experiments

DOE PAGES

Gomez, Matthew R.; Slutz, Stephen A.; Sefkow, Adam B.; ...

2015-04-29

In this study, the magnetized liner inertial fusion concept [S. A. Slutz et al., Phys. Plasmas17, 056303 (2010)] utilizes a magnetic field and laser heating to relax the pressure requirements of inertial confinement fusion. The first experiments to test the concept [M. R. Gomez et al., Phys. Rev. Lett. 113, 155003 (2014)] were conducted utilizing the 19 MA, 100 ns Z machine, the 2.5 kJ, 1 TW Z Beamlet laser, and the 10 T Applied B-field on Z system. Despite an estimated implosion velocity of only 70 km/s in these experiments, electron and ion temperatures at stagnation were as highmore » as 3 keV, and thermonuclear deuterium-deuterium neutron yields up to 2 × 10 12 have been produced. X-ray emission from the fuel at stagnation had widths ranging from 50 to 110 μm over a roughly 80% of the axial extent of the target (6–8 mm) and lasted approximately 2 ns. X-ray yields from these experiments are consistent with a stagnation density of the hot fuel equal to 0.2–0.4 g/cm 3. In these experiments, up to 5 ×10 10 secondary deuterium-tritium neutrons were produced. Given that the areal density of the plasma was approximately 1–2 mg/cm 2, this indicates the stagnation plasma was significantly magnetized, which is consistent with the anisotropy observed in the deuterium-tritium neutron spectra. Control experiments where the laser and/or magnetic field were not utilized failed to produce stagnation temperatures greater than 1 keV and primary deuterium-deuterium yields greater than 10 10. An additional control experiment where the fuel contained a sufficient dopant fraction to substantially increase radiative losses also failed to produce a relevant stagnation temperature. The results of these experiments are consistent with a thermonuclear neutron source.« less

Demonstration of thermonuclear conditions in magnetized liner inertial fusion experiments

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

Gomez, M. R.; Slutz, S. A.; Sefkow, A. B.

2015-05-15

The magnetized liner inertial fusion concept [S. A. Slutz et al., Phys. Plasmas 17, 056303 (2010)] utilizes a magnetic field and laser heating to relax the pressure requirements of inertial confinement fusion. The first experiments to test the concept [M. R. Gomez et al., Phys. Rev. Lett. 113, 155003 (2014)] were conducted utilizing the 19 MA, 100 ns Z machine, the 2.5 kJ, 1 TW Z Beamlet laser, and the 10 T Applied B-field on Z system. Despite an estimated implosion velocity of only 70 km/s in these experiments, electron and ion temperatures at stagnation were as high as 3 keV, and thermonuclear deuterium-deuterium neutronmore » yields up to 2 × 10{sup 12} have been produced. X-ray emission from the fuel at stagnation had widths ranging from 50 to 110 μm over a roughly 80% of the axial extent of the target (6–8 mm) and lasted approximately 2 ns. X-ray yields from these experiments are consistent with a stagnation density of the hot fuel equal to 0.2–0.4 g/cm{sup 3}. In these experiments, up to 5 × 10{sup 10} secondary deuterium-tritium neutrons were produced. Given that the areal density of the plasma was approximately 1–2 mg/cm{sup 2}, this indicates the stagnation plasma was significantly magnetized, which is consistent with the anisotropy observed in the deuterium-tritium neutron spectra. Control experiments where the laser and/or magnetic field were not utilized failed to produce stagnation temperatures greater than 1 keV and primary deuterium-deuterium yields greater than 10{sup 10}. An additional control experiment where the fuel contained a sufficient dopant fraction to substantially increase radiative losses also failed to produce a relevant stagnation temperature. The results of these experiments are consistent with a thermonuclear neutron source.« less

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

NASA Astrophysics Data System (ADS)

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

2018-06-01

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

Oscillations During Thermonuclear X-ray Bursts: A New Probe of Neutron Stars

NASA Technical Reports Server (NTRS)

Strohmayer, Tod E.; White, Nicholas E. (Technical Monitor)

2002-01-01

Observations of thermonuclear (also called Type 1) X-ray bursts from neutron stars in low mass X-ray binaries (LMXB) with the Rossi X-ray Timing Explorer (RXTE) have revealed large amplitude, high coherence X-ray brightness oscillations with frequencies in the 300 - 600 Hz range. Substantial spectral and timing evidence point to rotational modulation of the X-ray burst flux as the cause of these oscillations, and it is likely that they reveal the spin frequencies of neutron stars in LMXB from which they are detected. Here we review the status of our knowledge of these oscillations and describe how they can be used to constrain the masses and radii of neutron stars as well as the physics of thermonuclear burning on accreting neutron stars.

Development of the negative ion beams relevant to ITER and JT-60SA at Japan Atomic Energy Agency.

PubMed

Hanada, M; Kojima, A; Tobari, H; Nishikiori, R; Hiratsuka, J; Kashiwagi, M; Umeda, N; Yoshida, M; Ichikawa, M; Watanabe, K; Yamano, Y; Grisham, L R

2016-02-01

In order to realize negative ion sources and accelerators to be applicable to International Thermonuclear Experimental Reactor and JT-60 Super Advanced, a large cesium (Cs)-seeded negative ion source and a multi-aperture and multi-stage electric acceleration have been developed at Japan Atomic Energy Agency (JAEA). Long pulse production and acceleration of the negative ion beams have been independently carried out. The long pulse production of the high current beams has achieved 100 s at the beam current of 15 A by modifying the JT-60 negative ion source. The pulse duration time is increased three times longer than that before the modification. As for the acceleration, a pulse duration time has been also extended two orders of magnitudes from 0.4 s to 60 s. The developments of the negative ion source and acceleration at JAEA are well in progress towards the realization of the negative ion sources and accelerators for fusion applications.

Dense Plasma Focus: physics and applications (radiation material science, single-shot disclosure of hidden illegal objects, radiation biology and medicine, etc.)

NASA Astrophysics Data System (ADS)

Gribkov, V. A.; Miklaszewski, R.; Paduch, M.; Zielinska, E.; Chernyshova, M.; Pisarczyk, T.; Pimenov, V. N.; Demina, E. V.; Niemela, J.; Crespo, M.-L.; Cicuttin, A.; Tomaszewski, K.; Sadowski, M. J.; Skladnik-Sadowska, E.; Pytel, K.; Zawadka, A.; Giannini, G.; Longo, F.; Talab, A.; Ul'yanenko, S. E.

2015-03-01

The paper presents some outcomes obtained during the year of 2013 of the activity in the frame of the International Atomic Energy Agency Co-ordinated research project "Investigations of Materials under High Repetition and Intense Fusion-Relevant Pulses". The main results are related to the effects created at the interaction of powerful pulses of different types of radiation (soft and hard X-rays, hot plasma and fast ion streams, neutrons, etc. generated in Dense Plasma Focus (DPF) facilities) with various materials including those that are counted as perspective ones for their use in future thermonuclear reactors. Besides we discuss phenomena observed at the irradiation of biological test objects. We examine possible applications of nanosecond powerful pulses of neutrons to the aims of nuclear medicine and for disclosure of hidden illegal objects. Special attention is devoted to discussions of a possibility to create extremely large and enormously diminutive DPF devices and probabilities of their use in energetics, medicine and modern electronics.

Intermittent bursts induced by double tearing mode reconnection

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

Wei, Lai; Wang, Zheng-Xiong, E-mail: zxwang@dlut.edu.cn

Reversed magnetic shear (RMS) configuration is assumed to be the steady-state operation scenario for the future advanced tokamaks like International Thermonuclear Experimental Reactor. In this work, we numerically discover a phenomenon of violent intermittent bursts induced by self-organized double tearing mode (DTM) reconnection in the RMS configuration during the very long evolution, which may continuously lead to annular sawtooth crashes and thus badly impact the desired steady-state operation of the future advanced RMS tokamaks. The key process of the intermittent bursts in the off-axis region is similar to that of the typical sawtooth relaxation oscillation in the positive magnetic shearmore » configuration. It is interestingly found that in the decay phase of the DTM reconnection, the zonal field significantly counteracts equilibrium field to make the magnetic shear between the two rational surfaces so weak that the residual self-generated vortices of the previous DTM burst are able to trigger a reverse DTM reconnection by curling the field lines.« less

Compatibility of stainless steels and lithiated ceramics with beryllium

NASA Astrophysics Data System (ADS)

Flament, T.; Fauvet, P.; Sannier, J.

1988-07-01

The introduction of beryllium as a neutron multiplier in ceramic blankets of thermonuclear fusion reactors may give rise to the following compatibility problems: (i) oxidation of Be by ceramics (lithium aluminate and silicates) or by water vapour; (ii) interaction between beryllium and austenitic and martensitic steels. The studies were done in contact tests under vacuum and in tests under wet sweeping helium. The contact tests under vacuum have revealed that the interaction of beryllium with ceramics seems to be low up to 700°C, the interaction of beryllium with steels is significant and is characterized by the formation of a diffusion layer and of a brittle Be-Fe-Ni compound. With type 316 L austenitic steel, this interaction appears quite large at 600°C whereas it is noticeable only at 700°C with martensitic steels. The experiments carried out with sweeping wet helium at 600°C have evidenced a slight oxidation of beryllium due to water vapour which can be enhanced in the front of uncompletely dehydrated ceramics.

High heat flux burnout in subcooled flow boiling

NASA Astrophysics Data System (ADS)

Celata, G. P.; Cumo, M.; Mariani, A.

1995-09-01

The paper reports the results of an experimental research carried out at the Heat Transfer Division of the Energy Department, C.R. Casaccia, on the thermal hydraulic characterization of subcooled flow boiling CHF under typical conditions of thermonuclear fusion reactors, i.e. high liquid velocity and subcooling. The experiment was carried out exploring the following parameters: channel diameter (from 2.5 to 8.0 mm), heated length (10 and 15 cm), liquid velocity (from 2 to 40 m/s), exit pressure (from atmospheric to 5.0 MPa), inlet temperature (from 30 to 80 °C), channel orientation (vertical and horizontal). A maximum CHF value of 60.6 MW/m2 has been obtained under the following conditions: T in=30°, p=2.5 MPa, u=40 m/s, D=2.5 mm (smooth channel) Turbulence promoters (helically coiled wires) have been employed to further enhance the CHF attainable with subcooled flow boiling. Helically coiled wires allow an increase of 50% of the maximum CHF obtained with smooth channels.

Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

NASA Astrophysics Data System (ADS)

Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

2009-12-01

Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

Performance study of the neutron-TPC

NASA Astrophysics Data System (ADS)

Huang, Meng; Li, Yulan; Niu, Libo; Deng, Zhi; Cheng, Xiaolei; He, Li; Zhang, Hongyan; Fu, Jianqiang; Yan, Yangyang; Cai, Yiming; Li, Yuanjing

2017-02-01

Fast neutron spectrometers will play an important role in the future of the nuclear industry and nuclear physics experiments, in tasks such as fast neutron reactor monitoring, thermo-nuclear fusion plasma diagnostics, nuclear reaction cross-section measurement, and special nuclear material detection. Recently, a new fast neutron spectrometer based on a GEM (Gas Electron Multiplier amplification)-TPC (Time Projection Chamber), named the neutron-TPC, has been under development at Tsinghua University. It is designed to have a high energy resolution, high detection efficiency, easy access to the medium material, an outstanding n/γ suppression ratio, and a wide range of applications. This paper presents the design, test, and experimental study of the neutron-TPC. Based on the experimental results, the energy resolution (FWHM) of the neutron-TPC can reach 15.7%, 10.3% and 7.0% with detection efficiency higher than 10-5 for 1.2 MeV, 1.81 MeV and 2.5 MeV neutrons respectively. Supported by National Natural Science Foundation of China (11275109)

Intermittent bursts induced by double tearing mode reconnection

NASA Astrophysics Data System (ADS)

Wei, Lai; Wang, Zheng-Xiong

2014-06-01

Reversed magnetic shear (RMS) configuration is assumed to be the steady-state operation scenario for the future advanced tokamaks like International Thermonuclear Experimental Reactor. In this work, we numerically discover a phenomenon of violent intermittent bursts induced by self-organized double tearing mode (DTM) reconnection in the RMS configuration during the very long evolution, which may continuously lead to annular sawtooth crashes and thus badly impact the desired steady-state operation of the future advanced RMS tokamaks. The key process of the intermittent bursts in the off-axis region is similar to that of the typical sawtooth relaxation oscillation in the positive magnetic shear configuration. It is interestingly found that in the decay phase of the DTM reconnection, the zonal field significantly counteracts equilibrium field to make the magnetic shear between the two rational surfaces so weak that the residual self-generated vortices of the previous DTM burst are able to trigger a reverse DTM reconnection by curling the field lines.

Fracture toughness and fracture behavior of CLAM steel in the temperature range of 450 °C-550 °C

NASA Astrophysics Data System (ADS)

Zhao, Yanyun; Liang, Mengtian; Zhang, Zhenyu; Jiang, Man; Liu, Shaojun

2018-04-01

In order to analyze the fracture toughness and fracture behavior (J-R curves) of China Low Activation Martensitic (CLAM) steel under the design service temperature of Test Blanket Module of the International Thermonuclear Experimental Reactor, the quasi-static fracture experiment of CLAM steel was carried out under the temperature range of 450 °C-550 °C. The results indicated that the fracture behavior of CLAM steel was greatly influenced by test temperature. The fracture toughness increased slightly as the temperature increased from 450 °C to 500 °C. In the meanwhile, the fracture toughness at 550 °C could not be obtained due to the plastic deformation near the crack tip zone. The microstructure analysis based on the fracture topography and the interaction between dislocations and lath boundaries showed two different sub-crack propagation modes: growth along 45° of the main crack direction at 450 °C and growth perpendicular to the main crack at 500 °C.

Importance of helium-3 for the future

NASA Technical Reports Server (NTRS)

Kulcinski, Gerald L.

1989-01-01

Relevant plasma physics principles of thermonuclear research; the state of plasma physics as it pertains to the D-He(3) cycle; the technological benefits of the D-He(3) fuel cycle; the availability of He(3); and its location, methods of extraction and cost are discussed. A perspective on the rate of progress toward the goal of heating the confined plasma fuel to sufficiently high temperatures at high enough densities and for long enough times to cause substantial fusion of the atoms to take place is given in graphical form. The main technological advantages resulting from the D-He(3) fuel cycle, when compared with the DT cycle, are as follows: (1) increased electrical conversion efficiency; (2) reduced radiation damage to reactors; (3) reduced radioactive waste; (4) an increased level of safety in case of an accident; (5) the lower cost of electricity; and (6) the shorter time to commercialization. An account is given of mining He(3) on the Moon.

Frontier of Fusion Research: Path to the Steady State Fusion Reactor by Large Helical Device

NASA Astrophysics Data System (ADS)

Motojima, Osamu

2006-12-01

The ITER, the International Thermonuclear Experimental Reactor, which will be built in Cadarache in France, has finally started this year, 2006. Since the thermal energy produced by fusion reactions divided by the external heating power, i.e., the Q value, will be larger than 10, this is a big step of the fusion research for half a century trying to tame the nuclear fusion for the 6.5 Billion people on the Earth. The source of the Sun's power is lasting steadily and safely for 8 Billion years. As a potentially safe environmentally friendly and economically competitive energy source, fusion should provide a sustainable future energy supply for all mankind for ten thousands of years. At the frontier of fusion research important milestones are recently marked on a long road toward a true prototype fusion reactor. In its own merits, research into harnessing turbulent burning plasmas and thereby controlling fusion reaction, is one of the grand challenges of complex systems science. After a brief overview of a status of world fusion projects, a focus is given on fusion research at the National Institute for Fusion Science (NIFS) in Japan, which is playing a role of the Inter University Institute, the coordinating Center of Excellence for academic fusion research and by the Large Helical Device (LHD), the world's largest superconducting heliotron device, as a National Users' facility. The current status of LHD project is presented focusing on the experimental program and the recent achievements in basic parameters and in steady state operations. Since, its start in a year 1998, a remarkable progress has presently resulted in the temperature of 140 Million degree, the highest density of 500 Thousand Billion/cc with the internal density barrier (IDB) and the highest steady average beta of 4.5% in helical plasma devices and the largest total input energy of 1.6 GJ, in all magnetic confinement fusion devices. Finally, a perspective is given of the ITER Broad Approach program as an integrated part of ITER and Development of Fusion Energy project Agreement. Moreover, the relationship with the NIFS' new parent organization the National Institutes of Natural Sciences and with foreign research institutions is briefly explained.

Shock ignition of thermonuclear fuel with high areal density.

PubMed

Betti, R; Zhou, C D; Anderson, K S; Perkins, L J; Theobald, W; Solodov, A A

2007-04-13

A novel method by C. Zhou and R. Betti [Bull. Am. Phys. Soc. 50, 140 (2005)] to assemble and ignite thermonuclear fuel is presented. Massive cryogenic shells are first imploded by direct laser light with a low implosion velocity and on a low adiabat leading to fuel assemblies with large areal densities. The assembled fuel is ignited from a central hot spot heated by the collision of a spherically convergent ignitor shock and the return shock. The resulting fuel assembly features a hot-spot pressure greater than the surrounding dense fuel pressure. Such a nonisobaric assembly requires a lower energy threshold for ignition than the conventional isobaric one. The ignitor shock can be launched by a spike in the laser power or by particle beams. The thermonuclear gain can be significantly larger than in conventional isobaric ignition for equal driver energy.

A Game Theoretic Model of Thermonuclear Cyberwar

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

Soper, Braden C.

In this paper we propose a formal game theoretic model of thermonuclear cyberwar based on ideas found in [1] and [2]. Our intention is that such a game will act as a first step toward building more complete formal models of Cross-Domain Deterrence (CDD). We believe the proposed thermonuclear cyberwar game is an ideal place to start on such an endeavor because the game can be fashioned in a way that is closely related to the classical models of nuclear deterrence [4–6], but with obvious modifications that will help to elucidate the complexities introduced by a second domain. We startmore » with the classical bimatrix nuclear deterrence game based on the game of chicken, but introduce uncertainty via a left-of-launch cyber capability that one or both players may possess.« less

METHOD AND APPARATUS FOR CONTROLLING DIRECT-CYCLE NEUTRONIC REACTORS

DOEpatents

Reed, G.A.

1961-01-10

A control arrangement is offered for a boiling-water reactor. Boric acid is maintained in the water in the reactor and the amount in the reactor is controlled by continuously removing a portion of the water from the reactor, concentrating the boric acid by evaporating the water therefrom, returning a controlled amount of the acid to the reactor, and simultaneously controlling the water level by varying the rate of spent steam return to the reactor.

Frontiers in propulsion research: Laser, matter-antimatter, excited helium, energy exchange thermonuclear fusion

NASA Technical Reports Server (NTRS)

Papailiou, D. D. (Editor)

1975-01-01

Concepts are described that presently appear to have the potential for propulsion applications in the post-1990 era of space technology. The studies are still in progress, and only the current status of investigation is presented. The topics for possible propulsion application are lasers, nuclear fusion, matter-antimatter annihilation, electronically excited helium, energy exchange through the interaction of various fields, laser propagation, and thermonuclear fusion technology.

Thermonuclear runaways in nova outbursts

NASA Technical Reports Server (NTRS)

Shankar, Anurag; Arnett, David; Fryxell, Bruce A.

1992-01-01

Results of exploratory, two-dimensional numerical calculations of a local thermonuclear runaway on the surface of a white dwarf are reported. It is found that the energy released by the runaway can induce a significant amount of vorticity near the burning region. Such mass motions account naturally for mixing of core matter into the envelope during the explosion. A new mechanism for the lateral spread of nuclear burning is also discussed.

Publicly Released Prompt Radiation Spectra Suitable for Nuclear Detonation Simulations

DTIC Science & Technology

2017-03-01

emission. During the Hiroshima and Nagasaki bombings , the prompt radiation contributed from 40%-70% of the free-in-air dose depending on distance from...intermediate- and low-yield thermonuclear weapons for initial radiation shielding calculations No Gritzner, et al. 1976 ( EM -1, Low, Henre...DNA 4267F ( EM -1 Fission) Neutron Gritzner, et al. 1976 1.00 x 1023 Glasstone (Thermonuclear) Neutron Glasstone & Dolan 1977 1.445 x 1023 ORNL-TM

Fusion yield: Guderley model and Tsallis statistics

NASA Astrophysics Data System (ADS)

Haubold, H. J.; Kumar, D.

2011-02-01

The reaction rate probability integral is extended from Maxwell-Boltzmann approach to a more general approach by using the pathway model introduced by Mathai in 2005 (A pathway to matrix-variate gamma and normal densities. Linear Algebr. Appl. 396, 317-328). The extended thermonuclear reaction rate is obtained in the closed form via a Meijer's G-function and the so-obtained G-function is represented as a solution of a homogeneous linear differential equation. A physical model for the hydrodynamical process in a fusion plasma-compressed and laser-driven spherical shock wave is used for evaluating the fusion energy integral by integrating the extended thermonuclear reaction rate integral over the temperature. The result obtained is compared with the standard fusion yield obtained by Haubold and John in 1981 (Analytical representation of the thermonuclear reaction rate and fusion energy production in a spherical plasma shock wave. Plasma Phys. 23, 399-411). An interpretation for the pathway parameter is also given.

Results from MARBLE DT Experiments on the National Ignition Facility: Implosion of Foam-Filled Capsules for Studying Thermonuclear Burn in the Presence of Heterogeneous Mix

NASA Astrophysics Data System (ADS)

Murphy, T. J.; Douglas, M. R.; Cardenas, T.; Cooley, J. H.; Gunderson, M. A.; Haines, B. M.; Hamilton, C. E.; Kim, Y.; Lee, M. N.; Oertel, J. A.; Olson, R. E.; Randolph, R. B.; Shah, R. C.; Smidt, J. M.

2017-10-01

The MARBLE campaign on NIF investigates the effect of heterogeneous mix on thermonuclear burn for comparison to a probability distribution function (PDF) burn model. MARBLE utilizes plastic capsules filled with deuterated plastic foam and tritium gas. The ratio of DT to DD neutron yield is indicative of the degree to which the foam and the gas atomically mix. Platform development experiments have been performed to understand the behavior of the foam and of the gas separately using two types of capsule. The first experiments using deuterated foam and tritium gas have been performed. Results of these experiments, and the implications for our understanding of thermonuclear burn in heterogeneously mixed separated reactant experiments will be discussed. This work is supported by US DOE/NNSA, performed at LANL, operated by LANS LLC under contract DE-AC52-06NA25396.

On thermonuclear ignition criterion at the National Ignition Facility

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

Cheng, Baolian; Kwan, Thomas J. T.; Wang, Yi-Ming

2014-10-15

Sustained thermonuclear fusion at the National Ignition Facility remains elusive. Although recent experiments approached or exceeded the anticipated ignition thresholds, the nuclear performance of the laser-driven capsules was well below predictions in terms of energy and neutron production. Such discrepancies between expectations and reality motivate a reassessment of the physics of ignition. We have developed a predictive analytical model from fundamental physics principles. Based on the model, we obtained a general thermonuclear ignition criterion in terms of the areal density and temperature of the hot fuel. This newly derived ignition threshold and its alternative forms explicitly show the minimum requirementsmore » of the hot fuel pressure, mass, areal density, and burn fraction for achieving ignition. Comparison of our criterion with existing theories, simulations, and the experimental data shows that our ignition threshold is more stringent than those in the existing literature and that our results are consistent with the experiments.« less

Statistical methods for thermonuclear reaction rates and nucleosynthesis simulations

NASA Astrophysics Data System (ADS)

Iliadis, Christian; Longland, Richard; Coc, Alain; Timmes, F. X.; Champagne, Art E.

2015-03-01

Rigorous statistical methods for estimating thermonuclear reaction rates and nucleosynthesis are becoming increasingly established in nuclear astrophysics. The main challenge being faced is that experimental reaction rates are highly complex quantities derived from a multitude of different measured nuclear parameters (e.g., astrophysical S-factors, resonance energies and strengths, particle and γ-ray partial widths). We discuss the application of the Monte Carlo method to two distinct, but related, questions. First, given a set of measured nuclear parameters, how can one best estimate the resulting thermonuclear reaction rates and associated uncertainties? Second, given a set of appropriate reaction rates, how can one best estimate the abundances from nucleosynthesis (i.e., reaction network) calculations? The techniques described here provide probability density functions that can be used to derive statistically meaningful reaction rates and final abundances for any desired coverage probability. Examples are given for applications to s-process neutron sources, core-collapse supernovae, classical novae, and Big Bang nucleosynthesis.

Resistive edge mode instability in stellarator and tokamak geometries

NASA Astrophysics Data System (ADS)

Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

2008-09-01

Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

National Low-Level Waste Management Program Radionuclide Report Series. Volume 10, Nickel-63

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

Carboneau, M.L.; Adams, J.P.

1995-02-01

This report outlines the basic radiological, chemical, and physical characteristics of nickel-63 ({sup 63}Ni) and examines how these characteristics affect the behavior of {sup 63}Ni in various environmental media, such as soils, groundwater, plants, animals, the atmosphere, and the human body. Discussions also include methods of {sup 63}Ni production, waste types, and waste forms that contain {sup 63}Ni. The primary source of {sup 63}Ni in the environment has been low-level radioactive waste material generated as a result of neutron activation of stable {sup 62}Ni that is present in the structural components of nuclear reactor vessels. {sup 63}Ni enters the environmentmore » from the dismantling activities associated with nuclear reactor decommissioning. However, small amounts of {sup 63}Ni have been detected in the environment following the testing of thermonuclear weapons in the South Pacific. Concentrations as high as 2.7 Bq{sup a} per gram of sample (or equivalently 0.0022 parts per billion) were observed on Bikini Atoll (May 1954). {sup 63}Ni was not created as a fission product species (e.g., from {sup 235}U or {sup 239}Pu fissions), but instead was produced as a result of neutron capture in {sup 63}Ni, a common nickel isotope present in the stainless steel components of nuclear weapons (e.g., stainless-304 contains {approximately}9% total Ni or {approximately}0.3% {sup 63}Ni).« less

Impurity re-distribution in the corner regions of the JET divertor

NASA Astrophysics Data System (ADS)

Widdowson, A.; Coad, J. P.; Alves, E.; Baron-Wiechec, A.; Barradas, N. P.; Catarino, N.; Corregidor, V.; Heinola, K.; Krat, S.; Likonen, J.; Matthews, G. F.; Mayer, M.; Petersson, P.; Rubel, M.; Contributors, JET

2017-12-01

The International Thermonuclear Experimental Reactor (ITER) will use a mixture of deuterium (D) and tritium (T) as the fuel to generate power. Since T is both radioactive and expensive the Joint European Torus (JET) has been at the forefront of research to discover how much T is used and where it may be retained within the main reaction chamber. Until the year 2010 the JET plasma facing components were constructed of carbon fibre composites. During the JET carbon (C) phases impurities accumulated at the corners of the divertor located towards the bottom of the chamber in regions shadowed from the plasma where they are very difficult to reach and remove. This build-up of C and the associated H-isotope (including T) retention were of particular concern for future fusion reactors therefore, in 2010 JET changed the wall protection to (mainly) Be and the divertor to tungsten (W)—the JET ITER-like wall (ILW)—the choice of materials for ITER. This paper reveals that with the JET ILW impurities are still accumulating in the shadowed regions, with Be being the majority element, though the overall quantities are very much reduced from those in the C phases. Material will be transported into the shadowed regions principally when the plasma strike points are on the corner tiles, but particles typically have about a 75% probability of reflection from line-of sight surfaces, and multiple reflection/scattering results in deposition over all surfaces.

FROM THE HISTORY OF PHYSICS: The physics of a thermonuclear explosion of a normal-density liquefied deuterium sphere (On the impossibility of a spherically symmetric thermonuclear explosion in liquid deuterium at normal density)

NASA Astrophysics Data System (ADS)

Marchuk, Gurii I.; Imshennik, Vladimir S.; Basko, Mikhail M.

2009-03-01

The hydrodynamic problem of a thermonuclear explosion in a sphere of normal-density liquid deuterium was solved (Institute for Physics and Power Engineering, Obninsk) in 1952-1954 in the framework of the Soviet Atomic Project. The principal result was that the explosion shockwave in deuterium strongly decayed because of radiation energy loss and nonlocal energy release by fast neutrons. At that time, this negative result implied in essence that the straightforward approach to creating a thermonuclear weapon was in fact a blind alley. This paper describes a numerical solution to the stated problem, obtained with the modern DEIRA code developed for numerical modeling of inertially confined fusion. Detailed numerical calculations have confirmed the above 'historic' result and shed additional light on the physical causes of the detonation wave decay. The most pernicious factor is the radiation energy loss due to the combined effect of bremsstrahlung and the inverse Compton scattering of the emitted photons on the hot electrons. The impact of energy transfer by fast neutrons — which was already quite adequately accounted for in the above-cited historical work — is less significant. We present a more rigorous (compared to that of the 1950s) study of the role of inverse Compton scattering for which, in particular, an independent analytic estimate is obtained.

NUCLEAR REACTOR

DOEpatents

Treshow, M.

1961-09-01

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

Design of a Rail Gun System for Mitigating Disruptions in Fusion Reactors

NASA Astrophysics Data System (ADS)

Lay, Wei-Siang

Magnetic fusion devices, such as the tokamak, that carry a large amount of current to generate the plasma confining magnetic fields have the potential to lose magnetic stability control. This can lead to a major plasma disruption, which can cause most of the stored plasma energy to be lost to localized regions on the walls, causing severe damage. This is the most important issue for the $20B ITER device (International Thermonuclear Experimental Reactor) that is under construction in France. By injecting radiative materials deep into the plasma, the plasma energy could be dispersed more evenly on the vessel surface thus mitigating the harmful consequences of a disruption. Methods currently planned for ITER rely on the slow expansion of gases to propel the radiative payloads, and they also need to be located far away from the reactor vessel, which further slows down the response time of the system. Rail guns are being developed for aerospace applications, such as for mass transfer from the surface of the moon and asteroids to low earth orbit. A miniatured version of this aerospace technology seems to be particularly well suited to meet the fast time response needs of an ITER disruption mitigation system. Mounting this device close to the reactor vessel is also possible, which substantially increases its performance because the stray magnetic fields near the vessel walls could be used to augment the rail gun generated magnetic fields. In this thesis, the potential viability on Rail Gun based DMS is studied to investigate its projected fast time response capability by design, fabrication, and experiment of an NSTX-U sized rail gun system. Material and geometry based tests are used to find the most suitable armature design for this system for which the desirable attributes are high specific stiffness and high electrical conductivity. With the best material in these studies being aluminum 7075, the experimental Electromagnetic Particle Injector (EPI) system has propelled an aluminum armature (weighing 3g) to a velocity more than 150 m/s within two milliseconds post trigger, consistent with the predicted projection for a system with those parameters. Fixed magnetic field probes and high-speed images capture the velocity profile. To propel the armatures, a 20 mF capacitor bank charged to 2 kV and augmented with external field coils powers the rails. These studies indicate that an EPI based system can indeed operate with a fast response time of less than three milliseconds after an impending disruption is detected, and thus warrants further studies to more fully develop the concept as a back-up option for an ITER DMS.

Fast-acting nuclear reactor control device

DOEpatents

Kotlyar, Oleg M.; West, Phillip B.

1993-01-01

A fast-acting nuclear reactor control device for moving and positioning a fety control rod to desired positions within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump/motor, an electric gear motor, and solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

X-ray emission from high temperature plasmas

NASA Technical Reports Server (NTRS)

Harries, W. L.

1976-01-01

The physical processes occurring in plasma focus devices were studied. These devices produce dense high temperature plasmas, which emit X rays of hundreds of KeV energy and one to ten billion neutrons per pulse. The processes in the devices seem related to solar flare phenomena, and would also be of interest for controlled thermonuclear fusion applications. The high intensity, short duration bursts of X rays and neutrons could also possibly be used for pumping nuclear lasers.

Production of plasmas by long-wavelength lasers

DOEpatents

Dawson, J.M.

1973-10-01

A long-wavelength laser system for heating low-density plasma to high temperatures is described. In one embodiment, means are provided for repeatedly receiving and transmitting long-wavelength laser light in successive stages to form a laser-light beam path that repeatedly intersects with the equilibrium axis of a magnetically confined toroidal plasma column for interacting the laser light with the plasma for providing controlled thermonuclear fusion. Embodiments for heating specific linear plasmas are also provided. (Official Gazette)

α Heating in a Stagnated Z-pinch

NASA Astrophysics Data System (ADS)

Appelbe, Brian; Chittenden, Jeremy

2009-01-01

A computational investigation of a scheme for magneto-inertial confinement fusion in a Z-pinch is carried out. In the scheme implosion of a deuterium-tritium fuel mass is preceded by formation of a hotspot containing warm, dense plasma on axis. The presence of the hotspot increases energy yield. Compression of the hotspot by the main fuel mass initiates thermonuclear burn. There is significant heating of the plasma by thermonuclear α particles which are confined by the strong magnetic field of the Z-pinch.

The historical record for Sirius - Evidence for a white-dwarf thermonuclear runaway?

NASA Technical Reports Server (NTRS)

Bruhweiler, Frederick C.; Kondo, Yoji; Sion, Edward M.

1986-01-01

Evidence was recently presented that in medieval times Sirius was a bright red star, rather than the present bluish-white star. Here, the results of attempts to detect possible planetary nebula ejecta toward Sirius using data obtained by the IUE are presented. Based on these results and in the light of recent advances in understanding white-dwarf evolution, it is proposed that Sirius B underwent a recent thermonuclear runaway event triggered by a diffusion-induced CN reaction.

2009 EVALUATION OF TRITIUM REMOVAL AND MITIGATION TECHNOLOGIES FOR WASTEWATER TREATMENT

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

LUECK KJ; GENESSE DJ; STEGEN GE

2009-02-26

Since 1995, a state-approved land disposal site (SALDS) has received tritium contaminated effluents from the Hanford Site Effluent Treatment Facility (ETF). Tritium in this effluent is mitigated by storage in slow moving groundwater to allow extended time for decay before the water reaches the site boundary. By this method, tritium in the SALDS is isolated from the general environment and human contact until it has decayed to acceptable levels. This report contains the 2009 update evaluation of alternative tritium mitigation techniques to control tritium in liquid effluents and groundwater at the Hanford site. A thorough literature review was completed andmore » updated information is provided on state-of-the-art technologies for control of tritium in wastewaters. This report was prepared to satisfy the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-026-07B (Ecology, EPA, and DOE 2007). Tritium separation and isolation technologies are evaluated periodically to determine their feasibility for implementation to control Hanford site liquid effluents and groundwaters to meet the Us. Code of Federal Regulations (CFR), Title 40 CFR 141.16, drinking water maximum contaminant level (MCL) for tritium of 20,000 pOll and/or DOE Order 5400.5 as low as reasonably achievable (ALARA) policy. Since the 2004 evaluation, there have been a number of developments related to tritium separation and control with potential application in mitigating tritium contaminated wastewater. These are primarily focused in the areas of: (1) tritium recycling at a commercial facility in Cardiff, UK using integrated tritium separation technologies (water distillation, palladium membrane reactor, liquid phase catalytic exchange, thermal diffusion), (2) development and demonstration of Combined Electrolysis Catalytic Exchange (CECE) using hydrogen/water exchange to separate tritium from water, (3) evaporation of tritium contaminated water for dispersion in the atmosphere, and (4) use of barriers to minimize the transport of tritium in groundwater. Continuing development efforts for tritium separations processes are primarily to support the International Thermonuclear Experimental Reactor (ITER) program, the nuclear power industry, and the production of radiochemicals. While these applications are significantly different than the Hanford application, the technology could potentially be adapted for Hanford wastewater treatment. Separations based processes to reduce tritium levels below the drinking water MCL have not been demonstrated for the scale and conditions required for treating Hanford wastewater. In addition, available cost information indicates treatment costs for such processes will be substantially higher than for discharge to SALDS or other typical pump and treat projects at Hanford. Actual mitigation projects for groundwater with very low tritium contamination similar to that found at Hanford have focused mainly on controlling migration and on evaporation for dispersion in the atmosphere.« less

The Influence of Stellar Spin on Ignition of Thermonuclear Runaways

NASA Astrophysics Data System (ADS)

Galloway, Duncan K.; in ’t Zand, Jean J. M.; Chenevez, Jérôme; Keek, Laurens; Sanchez-Fernandez, Celia; Worpel, Hauke; Lampe, Nathanael; Kuulkers, Erik; Watts, Anna; Ootes, Laura; The MINBAR collaboration

2018-04-01

Runaway thermonuclear burning of a layer of accumulated fuel on the surface of a compact star provides a brief but intense display of stellar nuclear processes. For neutron stars accreting from a binary companion, these events manifest as thermonuclear (type-I) X-ray bursts, and recur on typical timescales of hours to days. We measured the burst rate as a function of accretion rate, from seven neutron stars with known spin rates, using a burst sample accumulated over several decades. At the highest accretion rates, the burst rate is lower for faster spinning stars. The observations imply that fast (>400 Hz) rotation encourages stabilization of nuclear burning, suggesting a dynamical dependence of nuclear ignition on the spin rate. This dependence is unexpected, because faster rotation entails less shear between the surrounding accretion disk and the star. Large-scale circulation in the fuel layer, leading to enhanced mixing of the burst ashes into the fuel layer, may explain this behavior; further numerical simulations are required to confirm this.

Development of real-time rotating waveplate Stokes polarimeter using multi-order retardation for ITER poloidal polarimeter

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

Imazawa, R., E-mail: imazawa.ryota@jaea.go.jp; Kawano, Y.; Ono, T.

The rotating waveplate Stokes polarimeter was developed for ITER (International Thermonuclear Experimental Reactor) poloidal polarimeter. The generalized model of the rotating waveplate Stokes polarimeter and the algorithm suitable for real-time field-programmable gate array (FPGA) processing were proposed. Since the generalized model takes into account each component associated with the rotation of the waveplate, the Stokes parameters can be accurately measured even in unideal condition such as non-uniformity of the waveplate retardation. Experiments using a He-Ne laser showed that the maximum error and the precision of the Stokes parameter were 3.5% and 1.2%, respectively. The rotation speed of waveplate was 20 000more » rpm and time resolution of measuring the Stokes parameter was 3.3 ms. Software emulation showed that the real-time measurement of the Stokes parameter with time resolution of less than 10 ms is possible by using several FPGA boards. Evaluation of measurement capability using a far-infrared laser which ITER poloidal polarimeter will use concluded that measurement error will be reduced by a factor of nine.« less

Development of real-time rotating waveplate Stokes polarimeter using multi-order retardation for ITER poloidal polarimeter.

PubMed

Imazawa, R; Kawano, Y; Ono, T; Itami, K

2016-01-01

The rotating waveplate Stokes polarimeter was developed for ITER (International Thermonuclear Experimental Reactor) poloidal polarimeter. The generalized model of the rotating waveplate Stokes polarimeter and the algorithm suitable for real-time field-programmable gate array (FPGA) processing were proposed. Since the generalized model takes into account each component associated with the rotation of the waveplate, the Stokes parameters can be accurately measured even in unideal condition such as non-uniformity of the waveplate retardation. Experiments using a He-Ne laser showed that the maximum error and the precision of the Stokes parameter were 3.5% and 1.2%, respectively. The rotation speed of waveplate was 20 000 rpm and time resolution of measuring the Stokes parameter was 3.3 ms. Software emulation showed that the real-time measurement of the Stokes parameter with time resolution of less than 10 ms is possible by using several FPGA boards. Evaluation of measurement capability using a far-infrared laser which ITER poloidal polarimeter will use concluded that measurement error will be reduced by a factor of nine.

Heat transfer in turbulent magneto-fluid-mechanic pipe flow

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

Andelman, M.P.

1975-12-01

The ability to predict heat transfer in Magneto-Fluid-Mechanic flow is of importance in light of the development of MHD generators and the proposed development of thermonuclear reactors. In both cases heat transfer from (or to) a conducting fluid in the presence of a magnetic field plays an important part in the overall economics of the system. A semi-empirical analytical method is given for obtaining heat transfer coefficients in turbulent liquid metal pipe flow in the presence of a magnetic field aligned to the flow. The analysis was based on the Lykoudis turbulent transport model with the influence of a longitudinalmore » magnetic field included. The results are shown to be in agreement with available experimental values. Experimental velocity profiles in mercury for pipe flow in a transverse magnetic field were made at a Reynolds number of 315,000; for Hartmann numbers of 0, 92, 184, 369, and 1198; and at orientations of 0 degrees, 45 degrees, and 90 degrees from the magnetic field. These results provide a basis for the determination of the effect of a transverse magnetic field on turbulent diffusivities.« less

A parametric study of helium retention in beryllium and its effect on deuterium retention

NASA Astrophysics Data System (ADS)

Alegre, D.; Baldwin, M. J.; Simmonds, M.; Nishijima, D.; Hollmann, E. M.; Brezinsek, S.; Doerner, R. P.

2017-12-01

Beryllium samples have been exposed in the PISCES-B linear plasma device to conditions relevant to the International Thermonuclear Experimental Reactor (ITER) in pure He, D, and D/He mixed plasmas. Except at intermediate sample exposure temperatures (573-673 K) He addition to a D plasma is found to have a beneficial effect as it reduces the D retention in Be (up to ˜55%), although the mechanism is unclear. Retention of He is typically around 1020-1021 He m-2, and is affected primarily by the Be surface temperature during exposition, by the ion fluence at <500 K exposure, but not by the ion impact energy at 573 K. Contamination of the Be surface with high-Z elements from the mask of the sample holder in pure He plasmas is also observed under certain conditions, and leads to unexpectedly large He retention values, as well as changes in the surface morphology. An estimation of the tritium retention in the Be first wall of ITER is provided, being sufficiently low to allow a safe operation of ITER.

Tokamak foundation in USSR/Russia 1950-1990

NASA Astrophysics Data System (ADS)

Smirnov, V. P.

2010-01-01

In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

Energy deposition and thermal effects of runaway electrons in ITER-FEAT plasma facing components

NASA Astrophysics Data System (ADS)

Maddaluno, G.; Maruccia, G.; Merola, M.; Rollet, S.

2003-03-01

The profile of energy deposited by runaway electrons (RAEs) of 10 or 50 MeV in International Thermonuclear Experimental Reactor-Fusion Energy Advanced Tokamak (ITER-FEAT) plasma facing components (PFCs) and the subsequent temperature pattern have been calculated by using the Monte Carlo code FLUKA and the finite element heat conduction code ANSYS. The RAE energy deposition density was assumed to be 50 MJ/m 2 and both 10 and 100 ms deposition times were considered. Five different configurations of PFCs were investigated: primary first wall armoured with Be, with and without protecting CFC poloidal limiters, both port limiter first wall options (Be flat tile and CFC monoblock), divertor baffle first wall, armoured with W. The analysis has outlined that for all the configurations but one (port limiter with Be flat tile) the heat sink and the cooling tube beneath the armour are well protected for both RAE energies and for both energy deposition times. On the other hand large melting (W, Be) or sublimation (C) of the surface layer occurs, eventually affecting the PFCs lifetime.

Simulation of the alpha particle heating and the helium ash source in an International Thermonuclear Experimental Reactor-like tokamak with an internal transport barrier

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

Ye, Lei, E-mail: lye@ipp.ac.cn; Guo, Wenfeng; Xiao, Xiaotao

2014-12-15

A guiding center orbit following code, which incorporates a set of non-singular coordinates for orbit integration, was developed and applied to investigate the alpha particle heating in an ITER-like tokamak with an internal transport barrier. It is found that a relatively large q (safety factor) value can significantly broaden the alpha heating profile in comparison with the local heating approximation; this broadening is due to the finite orbit width effects; when the orbit width is much smaller than the scale length of the alpha particle source profile, the heating profile agrees with the source profile, otherwise, the heating profile canmore » be significantly broadened. It is also found that the stagnation particles move to the magnetic axis during the slowing-down process, thus the effect of stagnation orbits is not beneficial to the helium ash removal. The source profile of helium ash is broadened in comparison with the alpha source profile, which is similar to the heating profile.« less

Compact NE213 neutron spectrometer with high energy resolution for fusion applications

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

Zimbal, A.; Reginatto, M.; Schuhmacher, H.

Neutron spectrometry is a tool for obtaining important information on the fuel ion composition, velocity distribution and temperature of fusion plasmas. A compact NE213 liquid scintillator, fully characterized at Physikalisch-Technische Bundesanstalt, was installed and operated at the Joint European Torus (JET) during two experimental campaigns (C8-2002 and trace tritium experiment-TTE 2003). The results show that this system can operate in a real fusion experiment as a neutron (1.5 MeV

Nuclear engine flow reactivity shim control

DOEpatents

Walsh, J.M.

1973-12-11

A nuclear engine control system is provided which automatically compensates for reactor reactivity uncertainties at the start of life and reactivity losses due to core corrosion during the reactor life in gas-cooled reactors. The coolant gas flow is varied automatically by means of specially provided control apparatus so that the reactor control drums maintain a predetermined steady state position throughout the reactor life. This permits the reactor to be designed for a constant drum position and results in a desirable, relatively flat temperature profile across the core. (Official Gazette)

Control rod drive for reactor shutdown

DOEpatents

McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

1976-01-20

A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

Development of tungsten armor and bonding to copper for plasma-interactive components

NASA Astrophysics Data System (ADS)

Smid, I.; Akiba, M.; Vieider, G.; Plöchl, L.

1998-10-01

For the highest sputtering threshold of all possible candidates, tungsten will be the most likely armor material in highly loaded plasma-interactive components of commercially relevant fusion reactors. The development of new materials, as well as joining and coating techniques are needed to find the best balance in plasma compatibility, lifetime, reliability, neutron irradiation resistance, and safety. Further important issues for selection are availability, costs of machining and production, etc. Tungsten doped with lanthanum oxide is a commercially available W grade for electrodes, designed for low electron work function, higher recrystallization temperature, reduced secondary grain growth, and machinability at relatively low costs. W-Re and related tungsten base alloys are preferred for application at high temperatures, when high strength, high thermal shock and recrystallization resistance are required. Due to the high costs and limited global availability of Re, however, the amount of such alloys in a commercial reactor should be kept low. Newly measured material properties up to high temperatures are presented for lanthanated and W-Re alloys, and the impact on fusion application is discussed. Recently developed coatings of chemical vapor deposited tungsten (CVD-W) on copper substrates have proven to be resistant to repeated thermal and shock loading. Layers of more than 5 mm, as required for the International Thermonuclear Experimental Reactor (ITER), became available. Vacuum plasma sprayed tungsten (VPS-W) in particular is attractive for its lower costs, and the potential of in situ repair. However, the advantage of sacrificial plasma-interactive tungsten coatings in long-term fusion devices has yet to be demonstrated. A durable and reliable joining of bulk tungsten to copper is needed to achieve an acceptable component lifetime in a fusion environment. The material properties of the copper alloys proposed for ITER, and their impact on the quality of bonding to tungsten is discussed. Future materials R&D should concern issues such as plasma compatibility, and above all neutron irradiation damage of promising tungsten-copper joints.

JEN-1 Reactor Control System; SISTEMA DE CONTROL DEL REACTOR JEN-1

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

Cantillo, M.F.; Nuno, C.M.; Andreu, J.L.M.

1963-01-01

ABS>The JEN-1 3Mw power swimming pool reactor electrical control circuits are described. Start-up, power generation in the core, and shutdown are controlled by the reactor control system. This control system guarantees in each moment the safety conditions during reactor operation. Each circuit was represented by a scheme, complemented with a description of its function, components, and operation theory. Components described include: scram circuit; fission counter control circuit; servo control circuit; control circuit of safety sheets; control circuits of primary, secondary, and clean-up pump motors and tower fan motor; primary valve motor circuit; center cubicle alarm circuit; and process alarm circuit.more » (auth)« less

Plasma control and utilization

DOEpatents

Ensley, Donald L.

1976-12-28

A plasma is confined and heated by a microwave field resonant in a cavity excited in a combination of the TE and TM modes while responding to the resonant frequency of the cavity as the plasma dimensions change to maintain operation at resonance. The microwave field is elliptically or circularly polarized as to prevent the electromagnetic confining field from going to zero. A high Q chamber having superconductive walls is employed to minimize wall losses while providing for extraction of thermonuclear energy produced by fusion of nuclei in the plasma.

Nuclear Coexistence: Rethinking U.S. Policy to Promote Stability in an Era of Proliferation

DTIC Science & Technology

1994-04-01

The Spread of Nuclear Weapons 1989 -90 (Boulder: Westview Press, 1990). 22. See William C. Martel and Steven E. Miller, "Controlling Borders and Nuclear...Security, Fall 1989 , Vol. 14, No. 2, pp. 140-41, for J. Robert Oppenheimer’s concerns about the dangers of the develop- ment of thermonuclear weapons. 4...Paradox of Technology," International Security, Vol. 14, No. 2 (Fall 1989 ), pp. 198-202. 6. Some argue that the U.S. strategy has relatively little

Nuclear reactor control column

DOEpatents

Bachovchin, Dennis M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

SPRING DRIVEN ACTUATING MECHANISM FOR NUCLEAR REACTOR CONTROL

DOEpatents

Bevilacqua, F.; Uecker, D.F.; Groh, E.F.

1962-01-23

l962. rod in a nuclear reactor to shut it down. The control rod or an extension thereof is wound on a drum as it is withdrawn from the reactor. When an emergency occurs requiring the reactor to be shut down, the drum is released so as to be free to rotate, and the tendency of the control rod or its extension coiled on the drum to straighten itself is used for quickly returning the control rod to the reactor. (AEC)

Thermionic switched self-actuating reactor shutdown system

DOEpatents

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

1989-01-01

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

Isotope separation by selective charge conversion and field deflection

DOEpatents

Hickman, Robert G.

1978-01-01

A deuterium-tritium separation system wherein a source beam comprised of positively ionized deuterium (D.sup.+) and tritium (T.sup.+) is converted at different charge-exchange cell sections of the system to negatively ionized deuterium (D.sup.-) and tritium (T.sup.-). First, energy is added to the beam to accelerate the D.sup.+ ions to the velocity that is optimum for conversion of the D.sup.+ ions to D.sup.- ions in a charge-exchange cell. The T.sup.+ ions are accelerated at the same time, but not to the optimum velocity since they are heavier than the D.sup.+ ions. The T.sup.+ ions are, therefore, not converted to T.sup.- ions when the D.sup.+ ions are converted to D.sup.- ions. This enables effective separation of the beam by deflection of the isotopes with an electrostatic field, the D.sup.- ions being deflected in one direction and the T.sup.+ ions being deflected in the opposite direction. Next, more energy is added to the deflected beam of T.sup.+ ions to bring the T.sup.+ ions to the optimum velocity for their conversion to T.sup.- ions. In a particular use of the invention, the beams of D.sup.- and T.sup.- ions are separately further accelerated and then converted to energetic neutral particles for injection as fuel into a thermonuclear reactor. The reactor exhaust of D.sup.+ and T.sup.+ and the D.sup.+ and T.sup.+ that was not converted in the respective sections is combined with the source beam and recycled through the system to increase the efficiency of the system.

Broadband sidebands generated by parametric instability in lower hybrid current drive experiments on EAST

NASA Astrophysics Data System (ADS)

Amicucci, L.; Ding, B. J.; Castaldo, C.; Cesario, R.; Giovannozzi, E.; Li, M. H.; Tuccillo, A. A.

2015-12-01

Modern research on nuclear fusion energy, based on the tokamak concept, has strong need of tools for actively driving non-inductive current especially at the periphery of plasma column, where tools available so far have poor efficiency. This is essential for solving one of the most critical problems for thermonuclear reactor, consisting in how to achieve the figure of fusion gain in the context of sufficient stability. The lower hybrid current drive (LHCD) effect has the potential capability of driving current at large radii of reactor plasma with high efficiency [1]. Experiments recently carried out on EAST showed that a strong activity of LH sideband waves (from the RF probe spectra), accompanied by weak core penetration of the coupled LH power, is present when operating at relatively high plasma densities. Previous theoretical results, confirmed by experiments on FTU, showed that the LH sideband phenomenon is produced by parametric instability (PI), which are mitigated by higher plasma edge temperatures. This condition is thus useful for enabling the LH power propagation when operating with profiles having high plasma densities even at the edge. In the present work, we show new PI modeling of EAST plasmas data, obtained in condition of higher plasma edge temperature due to chamber lithisation. The obtained trend of the PI frequencies and growth rates is consistent with data of RF probe spectra, available in different regimes of lithisated and not lithisated vessel. Moreover, these spectra are interpreted as PI effect occurring at the periphery of plasma column, however in the low field side where the LH power is coupled.

Robustness of waves with a high phase velocity

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

Tajima, T., E-mail: ttajima@uci.edu; Tri Alpha Energy, Inc., P.O. Box 7010, Rancho Santa Margarita, CA 92688; Necas, A., E-mail: anecas@trialphaenergy.com

Norman Rostoker pioneered research of (1) plasma-driven accelerators and (2) beam-driven fusion reactors. The collective acceleration, coined by Veksler, advocates to drive above-ionization plasma waves by an electron beam to accelerate ions. The research on this, among others, by the Rostoker group incubated the idea that eventually led to the birth of the laser wakefield acceleration (LWFA), by which a large and robust accelerating collective fields may be generated in plasma in which plasma remains robust and undisrupted. Besides the emergence of LWFA, the Rostoker research spawned our lessons learned on the importance of adiabatic acceleration of ions in collectivemore » accelerators, including the recent rebirth in laser-driven ion acceleration efforts in a smooth adiabatic fashion by a variety of ingenious methods. Following Rostoker’s research in (2), the beam-driven Field Reversed Configuration (FRC) has accomplished breakthroughs in recent years. The beam-driven kinetic plasma instabilities have been found to drive the reactivity of deuteron-deuteron fusion beyond the thermonuclear yield in C-2U plasma that Rostoker started. This remarkable result in FRCs as well as the above mentioned LWFA may be understood with the aid of the newly introduced idea of the “robustness hypothesis of waves with a high phase velocity”. It posits that when the wave driven by a particle beam (or laser pulse) has a high phase velocity, its amplitude is high without disrupting the supporting bulk plasma. This hypothesis may guide us into more robust and efficient fusion reactors and more compact accelerators.« less

NUCLEAR REACTOR CONTROL SYSTEM

DOEpatents

Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

1959-11-01

A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

Real-Time Processing System for the JET Hard X-Ray and Gamma-Ray Profile Monitor Enhancement

NASA Astrophysics Data System (ADS)

Fernandes, Ana M.; Pereira, Rita C.; Neto, André; Valcárcel, Daniel F.; Alves, Diogo; Sousa, Jorge; Carvalho, Bernardo B.; Kiptily, Vasily; Syme, Brian; Blanchard, Patrick; Murari, Andrea; Correia, Carlos M. B. A.; Varandas, Carlos A. F.; Gonçalves, Bruno

2014-06-01

The Joint European Torus (JET) is currently undertaking an enhancement program which includes tests of relevant diagnostics with real-time processing capabilities for the International Thermonuclear Experimental Reactor (ITER). Accordingly, a new real-time processing system was developed and installed at JET for the gamma-ray and hard X-ray profile monitor diagnostic. The new system is connected to 19 CsI(Tl) photodiodes in order to obtain the line-integrated profiles of the gamma-ray and hard X-ray emissions. Moreover, it was designed to overcome the former data acquisition (DAQ) limitations while exploiting the required real-time features. The new DAQ hardware, based on the Advanced Telecommunication Computer Architecture (ATCA) standard, includes reconfigurable digitizer modules with embedded field-programmable gate array (FPGA) devices capable of acquiring and simultaneously processing data in real-time from the 19 detectors. A suitable algorithm was developed and implemented in the FPGAs, which are able to deliver the corresponding energy of the acquired pulses. The processed data is sent periodically, during the discharge, through the JET real-time network and stored in the JET scientific databases at the end of the pulse. The interface between the ATCA digitizers, the JET control and data acquisition system (CODAS), and the JET real-time network is provided by the Multithreaded Application Real-Time executor (MARTe). The work developed allowed attaining two of the major milestones required by next fusion devices: the ability to process and simultaneously supply high volume data rates in real-time.

Control of autothermal reforming reactor of diesel fuel

NASA Astrophysics Data System (ADS)

Dolanc, Gregor; Pregelj, Boštjan; Petrovčič, Janko; Pasel, Joachim; Kolb, Gunther

2016-05-01

In this paper a control system for autothermal reforming reactor for diesel fuel is presented. Autothermal reforming reactors and the pertaining purification reactors are used to convert diesel fuel into hydrogen-rich reformate gas, which is then converted into electricity by the fuel cell. The purpose of the presented control system is to control the hydrogen production rate and the temperature of the autothermal reforming reactor. The system is designed in such a way that the two control loops do not interact, which is required for stable operation of the fuel cell. The presented control system is a part of the complete control system of the diesel fuel cell auxiliary power unit (APU).

Bioaugmentation of activated sludge towards 3-chloroaniline removal with a mixed bacterial population carrying a degradative plasmid.

PubMed

Bathe, Stephan; Schwarzenbeck, Norbert; Hausner, Martina

2009-06-01

A bioaugmentation approach combining several strategies was applied to achieve degradation of 3-chloroaniline (3CA) in semicontinuous activated sludge reactors. In a first step, a 3CA-degrading Comamonas testosteroni strain carrying the degradative plasmid pNB2 was added to a biofilm reactor, and complete 3CA degradation together with spread of the plasmid within the indigenous biofilm population was achieved. A second set of reactors was then bioaugmented with either a suspension of biofilm cells removed from the carrier material or with biofilm-containing carrier material. 3CA degradation was established rapidly in all bioaugmented reactors, followed by a slow adaptation of the non-bioaugmented control reactors. In response to variations in 3CA concentration, all reactors exhibited temporary performance breakdowns. Whereas duplicates of the control reactors deviated in their behaviour, the bioaugmented reactors appeared more reproducible in their performance and population dynamics. Finally, the carrier-bioaugmented reactors showed an improved performance in the presence of high 3CA influent concentrations over the suspension-bioaugmented reactors. In contrast, degradation in one control reactor failed completely, but was rapidly established in the remaining control reactor.

Reactivity control assembly for nuclear reactor. [LMFBR

DOEpatents

Bollinger, L.R.

1982-03-17

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

NUCLEAR REACTOR AS THE OBJECT OF CONTROL. AUTOMATIC CONTROL OF AIRCRAFT ENGINES . B.S. Voronkev Collection of Articles

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

None

BS> The dynamics of a power reactor is treated in some detail. Although the reactor is described by a nonlinear differential equation of the seventh order, a two-group approximstion with prompt neutrons and one averaged group of delayed neutrons may be used. When the reactor is in equilibrium, the reactor equation may be linearized in two ways. The effects of positive and negative coefficients of tins of the reactor are discussed. The nonlinear character of the control rods is trested. (D.L.C.)

30S(α , p) Thermonuclear Reaction Rate from Experimental Level Structure of 34Ar

NASA Astrophysics Data System (ADS)

Kahl, D.; Chen, A. A.; Kubono, S.; Yamaguchi, H.; Binh, D. N.; Chen, J.; Cherubini, S.; Duy, N. N.; Hashimoto, T.; Hayakawa, S.; Iwasa, N.; Jung, H. S.; Kato, S.; Kwon, Y. K.; Nishimura, S.; Ota, S.; Setoodehnia, K.; Teranishi, T.; Tokieda, H.; Yamada, T.; Yun, C. C.; Zhang, L. Y.

Type I X-ray bursts are the most frequent thermonuclear explosions in the galaxy. Owing to their recurrence from known astronomical objects, burst morphology is extensively documented, and they are modeled very successfully as neutron-deficient, thermonuclear runaway on the surface of accreting neutron stars. While reaction networks include hundreds of isotopes and thousands of nuclear processes, only a small subset appear to play a pivotal role. One such reaction is the 30S(α , p) reaction, which is believed to be a crucial link in the explosive helium burning which is responsible for the large energy flux. However, very little experimental information is available concerning the cross section itself, nor the 34Ar compound nucleus at the relevant energies. We performed the first study of the entrance channel via 30S alpha resonant elastic scattering using a state-of-the-art, low-energy, 30S radioactive ion beam. The measurement was performed in inverse kinematics using a newly-developed active target. An R-matrix analysis of the excitation function reveals previously unknown resonances, including their quantum properties of spin, parity, width, and energy.

Thermonuclear reactions probed at stellar-core conditions with laser-based inertial-confinement fusion

DOE PAGES

Casey, D. T.; Sayre, D. B.; Brune, C. R.; ...

2017-08-07

Stars are giant thermonuclear plasma furnaces that slowly fuse the lighter elements in the universe into heavier elements, releasing energy, and generating the pressure required to prevent collapse. To understand stars, we must rely on nuclear reaction rate data obtained, up to now, under conditions very different from those of stellar cores. Here we show thermonuclear measurements of the 2H(d, n) 3He and 3H(t,2n) 4He S-factors at a range of densities (1.2–16 g cm –3) and temperatures (2.1–5.4 keV) that allow us to test the conditions of the hydrogen-burning phase of main-sequence stars. The relevant conditions are created using inertial-confinementmore » fusion implosions at the National Ignition Facility. Here, our data agree within uncertainty with previous accelerator-based measurements and establish this approach for future experiments to measure other reactions and to test plasma-nuclear effects present in stellar interiors, such as plasma electron screening, directly in the environments where they occur.« less

Results from the MARBLE Campaign on the National Ignition Facility: Implosion of Foam-Filled Capsules for Studying Thermonuclear Burn in the Presence of Heterogeneous Mix

NASA Astrophysics Data System (ADS)

Murphy, T. J.; Douglas, M. R.; Cardenas, T.; Devolder, B. G.; Fincke, J. R.; Gunderson, M. A.; Haines, B. M.; Hamilton, C. E.; Kim, Y. H.; Lee, M. N.; Oertel, J. A.; Olson, R. E.; Randolph, R. B.; Shah, R. C.; Smidt, J. M.

2016-10-01

The MARBLE campaign on NIF investigates the effect of heterogeneous mix on thermonuclear burn for comparison to a probability distribution function (PDF) burn model. MARBLE utilizes plastic capsules filled with deuterated plastic foam and tritium gas. The ratio of DT to DD neutron yield is indicative of the degree to which the foam and the gas atomically mix. Platform development experiments have been performed to understand the behavior of the foam and of the gas separately using two types of capsule. The first uses partially deuterated foam and hydrogen gas fill to understand the burn in the foam. The second uses undeuterated foam and deuterium gas fill to understand the dynamics of the gas. Experiments using deuterated foam and tritium gas are planned. Results of these experiments, and the implications for our understanding of thermonuclear burn in heterogeneously mixed separated reactant experiments will be discussed. This work is supported by US DOE/NNSA, performed at LANL, operated by LANS LLC under contract DE-AC52-06NA25396.

The LOFT perspective on neutron star thermonuclear bursts: White paper in support of the mission concept of the large observatory for X-ray timing

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

in't Zand, J. J.M.; Malone, Christopher M.; Altamirano, D.

2015-01-14

The Large Area Detector (LAD) on the Large Observatory For X-ray Timing ( LOFT ), with a 8.5 m 2 photon- collecting area in the 2–30 keV bandpass at CCD-class spectral resolving power (λ/Δλ = 10 – 100), is designed for optimum performance on bright X-ray sources. Thus, it is well-suited to study thermonuclear X-ray bursts from Galactic neutron stars. These bursts will typically yield 2 x 10 5 photon detections per second in the LAD, which is at least 15 times more than with any other instrument past, current or anticipated. The Wide Field Monitor (WFM) foreseen for LOFTmore » uniquely combines 2–50 keV imaging with large (30%) prompt sky coverage. This will enable the detection of tens of thousands of thermonuclear X-ray bursts during a 3-yr mission, including tens of superbursts. Both numbers are similar or more than the current database gathered in 50 years of X-ray astronomy.« less

Autonomous Control of Space Nuclear Reactors

NASA Technical Reports Server (NTRS)

Merk, John

2013-01-01

Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

The Complete Burning of Weapons Grade Plutonium and Highly Enriched Uranium with (Laser Inertial Fusion-Fission Energy) LIFE Engine

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

Farmer, J C; Diaz de la Rubia, T; Moses, E

2008-12-23

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spentmore » nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including unenriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to divert in large quantities. This report discusses the application of the LIFE concept to nonproliferation issues, initially looking at the LIFE (Laser Inertial Fusion-Fission Energy) engine as a means of completely burning WG Pu and HEU. By combining a neutron-rich inertial fusion point source with energy-rich fission, the once-through closed fuel-cycle LIFE concept has the following characteristics: it is capable of efficiently burning excess weapons or separated civilian plutonium and highly enriched uranium; the fission blanket is sub-critical at all times (keff < 0.95); because LIFE can operate well beyond the point at which light water reactors (LWRs) need to be refueled due to burn-up of fissile material and the resulting drop in system reactivity, fuel burn-up of 99% or more appears feasible. The objective of this work is to develop LIFE technology for burning of WG-Pu and HEU.« less

Control Means for Reactor

DOEpatents

Manley, J. H.

1961-06-27

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Thermonuclear runaways in thick hydrogen rich envelopes of neutron stars

NASA Technical Reports Server (NTRS)

Starrfield, S. G.; Kenyon, S.; Truran, J. W.; Sparks, W. M.

1981-01-01

A Lagrangian, fully implicit, one dimensional hydrodynamic computer code was used to evolve thermonuclear runaways in the accreted hydrogen rich envelopes of 1.0 Msub solar neutron stars with radii of 10 km and 20 km. Simulations produce outbursts which last from about 750 seconds to about one week. Peak effective temeratures and luninosities were 26 million K and 80 thousand Lsub solar for the 10 km study and 5.3 millison and 600 Lsub solar for the 20 km study. Hydrodynamic expansion on the 10 km neutron star produced a precursor lasting about one ten thousandth seconds.

Burst Oscillations: A New Spin on Neutron Stars

NASA Technical Reports Server (NTRS)

Strohmayer, Tod

2007-01-01

Observations with NASA's Rossi X-ray Timing Explorer (RXTE) have shown that the X-ray flux during thermonuclear X-ray bursts fr-om accreting neutron stars is often strongly pulsed at frequencies as high as 620 Hz. We now know that these oscillations are produced by spin modulation of the thermonuclear flux from the neutron star surface. In addition to revealing the spin frequency, they provide new ways to probe the properties and physics of accreting neutron stars. I will briefly review our current observational and theoretical understanding of these oscillations and discuss what they are telling us about neutron stars.

Reactivity control assembly for nuclear reactor

DOEpatents

Bollinger, Lawrence R.

1984-01-01

Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

Operators in the Plum Brook Reactor Facility Control Room

NASA Image and Video Library

1970-03-21

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

ELECTRICAL COIL STRUCTURE

DOEpatents

Baker, W.R.; Hartwig, A.

1962-09-25

A compactly wound electrical coil is designed for carrying intense pulsed currents such as are characteristic of controlled thermonuclear reaction devices. A flat strip of conductor is tightly wound in a spiral with a matching flat strip of insulator. To provide for a high fluid coolant flow through the coil with minimum pumping pressure, a surface of the conductor is scored with parallel transverse grooves which form short longitudinal coolant pasaages when the conductor is wound in the spiral configuration. Owing to this construction, the coil is extremely resistant to thermal and magnetic shock from sudden high currents. (AEC)

TREAT Reactor Control and Protection System

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

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS).more » The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab.« less

77 FR 41206 - Guidelines for Preparing and Reviewing Licensing Applications for Instrumentation and Control...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-12

... Applications for Instrumentation and Control Upgrades for Non-Power Reactors AGENCY: Nuclear Regulatory... (NRC or the Commission) is requesting public comment on Chapter 7, Section 7.3, Reactor Control System...-Power Reactors: Format and Content,'' for instrumentation and control (I&C) upgrades and NUREG-1537...

Fail-safe reactivity compensation method for a nuclear reactor

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

Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on themore » constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.« less

APPARATUS FOR THE DENSIFICATION AND ENERGIZATION OF CHARGED PARTICLES

DOEpatents

Post, R.F.; Coensgen, F.H.

1962-12-18

This patent relates to a device for materially increasing the energy and density of a plasma to produce conditions commensurate with the establishment and promotion of controlled thermonuclear reactions. To this end the device employs three successive stages of magnetic compression, each stage having magnetic mirrors to compress a plasma, the mirrors being moveable to transfer the plasma to successive stages for further compression. Accordingly, a plasma introduced to the first stage is increased in density and energy in stepwide fashion by virtue of the magnetic compression in the successive stages such that the plasma upon reaching the last stage is of extremely high energy and density commensurate the plasma particles undergoing thermonuclear reactions. The principal novelty of the device resides in the provision of a unidirectional magnetic field which increases in stepwise fashion in coaxially communicating compression chambers of progressively decreasing lengths and diameters. Pulsed magnetic fields are superimposed upon the undirectional field and are manipulated to establish resultant magnetic compression fields which increase in intensity and progressively move, with respect to time, through the compression chambers in the direction of the smallest one thereof. The resultant field in the last compression chamber is hence of relatively high intensity, and the density and energy of the plasma confined therein are correspondingly high. (AEC)

Review of controlled fusion research using laser heating.

NASA Technical Reports Server (NTRS)

Hertzberg, A.

1973-01-01

Development of methods for generating high laser pulse energy has stimulated research leading to new ideas for practical controlled thermonuclear fusion machines. A review is presented of some important efforts in progress, and two different approaches have been selected as examples for discussion. One involves the concept of very short pulse lasers with power output tailored, in time, to obtain a nearly isentropic compression of a deuterium-tritium pellet to very high densities and temperatures. A second approach utilizing long wavelength, long pulse, efficient gas lasers to heat a column of plasma contained in a solenoidal field is also discussed. The working requirements of the laser and various magnetic field geometries of this approach are described.

System and method for air temperature control in an oxygen transport membrane based reactor

DOEpatents

Kelly, Sean M

2016-09-27

A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

System and method for temperature control in an oxygen transport membrane based reactor

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

Kelly, Sean M.

A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

CONTROL FOR NEUTRONIC REACTOR

DOEpatents

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

1959-03-31

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

BOILING SLURRY REACTOR AND METHOD FO CONTROL

DOEpatents

Petrick, M.; Marchaterre, J.F.

1963-05-01

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

Dual annular rotating "windowed" nuclear reflector reactor control system

DOEpatents

Jacox, Michael G.; Drexler, Robert L.; Hunt, Robert N. M.; Lake, James A.

1994-01-01

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core.

Self-teaching neural network learns difficult reactor control problem

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

Jouse, W.C.

1989-01-01

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

NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

DOEpatents

Young, G.J.

1959-06-30

The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

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

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

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

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

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

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Combustible gas control for nuclear power reactors. 50.44... FACILITIES Standards for Licenses, Certifications, and Regulatory Approvals § 50.44 Combustible gas control... capability for ensuring a mixed atmosphere. (2) Combustible gas control. (i) All boiling water reactors with...

Dynamic Modeling and Control of Nuclear Reactors Coupled to Closed-Loop Brayton Cycle Systems using SIMULINK{sup TM}

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

Wright, Steven A.; Sanchez, Travis

2005-02-06

The operation of space reactors for both in-space and planetary operations will require unprecedented levels of autonomy and control. Development of these autonomous control systems will require dynamic system models, effective control methodologies, and autonomous control logic. This paper briefly describes the results of reactor, power-conversion, and control models that are implemented in SIMULINK{sup TM} (Simulink, 2004). SIMULINK{sup TM} is a development environment packaged with MatLab{sup TM} (MatLab, 2004) that allows the creation of dynamic state flow models. Simulation modules for liquid metal, gas cooled reactors, and electrically heated systems have been developed, as have modules for dynamic power-conversion componentsmore » such as, ducting, heat exchangers, turbines, compressors, permanent magnet alternators, and load resistors. Various control modules for the reactor and the power-conversion shaft speed have also been developed and simulated. The modules are compiled into libraries and can be easily connected in different ways to explore the operational space of a number of potential reactor, power-conversion system configurations, and control approaches. The modularity and variability of these SIMULINK{sup TM} models provides a way to simulate a variety of complete power generation systems. To date, both Liquid Metal Reactors (LMR), Gas Cooled Reactors (GCR), and electric heaters that are coupled to gas-dynamics systems and thermoelectric systems have been simulated and are used to understand the behavior of these systems. Current efforts are focused on improving the fidelity of the existing SIMULINK{sup TM} modules, extending them to include isotopic heaters, heat pipes, Stirling engines, and on developing state flow logic to provide intelligent autonomy. The simulation code is called RPC-SIM (Reactor Power and Control-Simulator)« less

Control system for a small fission reactor

DOEpatents

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

1986-01-01

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

Armour Materials for the ITER Plasma Facing Components

NASA Astrophysics Data System (ADS)

Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

Development of the Long Pulse Negative Ion Source for ITER

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

Hemsworth, R.S.; Svensson, L.; Esch, H.P.L. de

2005-04-06

A model of the ion source designed for the neutral beam injectors of the International Thermonuclear Experimental Reactor (ITER), the KAMABOKO III ion source, is being tested on the MANTIS test stand at the DRFC Cadarache in collaboration with JAERI, Japan, who designed and supplied the ion source. The ion source is attached to a 3 grid 30 keV accelerator (also supplied by JAERI) and the accelerated negative ion current is determined from the energy deposited on a calorimeter located 1.6 m from the source.During experiments on MANTIS three adverse effects of long pulse operation were found: The negative ionmore » current to the calorimeter is {approx_equal}50% of that obtained from short pulse operation Increasing the plasma grid (PG) temperature results in {<=}40% enhancement in negative ion yield, substantially below that reported for short pulse operation, {>=}100%. The caesium 'consumption' is up to 1500 times that expected.Results presented here indicate that each of these is, at least partially, explained by thermal effects. Additionally presented are the results of a detailed characterisation of the source, which enable the most efficient mode of operation to be identified.« less

Gas Evolution from Insulating Materials for Superconducting Coil of Iter by Gamma Ray Irradiation at Liquid Nitrogen Temperature

NASA Astrophysics Data System (ADS)

Idesaki, A.; Koizumi, N.; Sugimoto, M.; Morishita, N.; Ohshima, T.; Okuno, K.

2008-03-01

A laminated material composed of glass cloth/polyimide film/epoxy resin will be used as an insulating material for superconducting coil of International Thermonuclear Experimental Reactor (ITER). In order to keep safe and stable operation of the superconducting coil system, it is indispensable to evaluate radiation resistance of the material, because the material is exposed to severe environments such as high radiation field and low temperature of 4 K. Especially, it is important to estimate the amount of gases evolved from the insulating material by irradiation, because the gases affect on the purifying system of liquid helium in the superconducting coil system. In this work, the gas evolution from the laminated material by gamma ray irradiation at liquid nitrogen temperature (77 K) was investigated, and the difference of gas evolution behavior due to difference of composition in the epoxy resin was discussed. It was found that the main gases evolved from the laminated material by the irradiation were hydrogen, carbon monoxide and carbon dioxide, and that the amount of gases evolved from the epoxy resin containing cyanate ester was about 60% less than that from the epoxy resin containing tetraglycidyl-diaminophenylmethane (TGDDM).

Synergistic effects of surface erosion on tritium inventory and permeation in metallic plasma facing armours

NASA Astrophysics Data System (ADS)

Federici, G.; Holland, D. F.; Matera, R.

1996-10-01

In the next generation of DT fuelled tokamaks, i.e., the International Thermonuclear Experimental Reactor (ITER) implantation of energetic DT particles on some portions of the plasma facing components (PFCs) will take place along with significant erosion of the armour surfaces. As a result of the simultaneous removal of material from the front surface, the build-up of tritium inventory and the start of permeation originating in the presence of large densities of neutron-induced traps is expected to be influenced considerably and special provisions could be required to minimise the consequences on the design. This paper reports on the results of a tritium transport modelling study based on a new model which describes the migration of implanted tritium across the bulk of metallic plasma facing materials containing neutron-induced traps which can capture it and includes the synergistic effects of surface erosion. The physical basis of the model is summarised, but emphasis is on the discussion of the results of a comparative study performed for beryllium and tungsten armours for ranges of design and operation conditions similar to those anticipated in the divertor of ITER.

Current Drive

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

Faulconer, D.W

2004-03-15

Certain devices aimed at magnetic confinement of thermonuclear plasma rely on the steady flow of an electric current in the plasma. In view of the dominant place it occupies in both the world magnetic-confinement fusion effort and the author's own activity, the tokamak toroidal configuration is selected as prototype for discussing the question of how such a current can be maintained. Tokamaks require a stationary toroidal plasma current, this being traditionally provided by a pulsed magnetic induction which drives the plasma ring as the secondary of a transformer. Since this mechanism is essentially transient, and steady-state fusion reactor operation hasmore » manifold advantages, significant effort is now devoted to developing alternate steady-state means of generating toroidal current. These methods are classed under the global heading of 'noninductive current drive' or simply 'current drive', generally, though not exclusively, employing the injection of waves and/or toroidally directed particle beams. In what follows we highlight the physical mechanisms underlying surprisingly various approaches to driving current in a tokamak, downplaying a number of practical and technical issues. When a significant data base exists for a given method, its experimental current drive efficiency and future prospects are detailed.« less

Short initial length quench on CICC of ITER TF coils

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

Nicollet, S.; Ciazynski, D.; Duchateau, J.-L.

Previous quench studies performed for the International Thermonuclear Experimental Reactor (ITER) Toroidal Field (TF) Coils have led to identify two extreme families of quench: first 'severe' quenches over long initial lengths in high magnetic field, and second smooth quenches over short initial lengths in low field region. Detailed analyses and results on smooth quench propagation and detectability on one TF Cable In Conduit Conductor (CICC) with a lower propagation velocity are presented here. The influence of the initial quench energy is shown and results of computations with either a Fast Discharge (FD) of the magnet or without (failure of themore » voltage quench detection system) are reported. The influence of the central spiral of the conductor on the propagation velocity is also detailed. In the cases of a regularly triggered FD, the hot spot temperature criterion of 150 K (with helium and jacket) is fulfilled for an initial quench length of 1 m, whereas this criterion is exceed (Tmax ≈ 200 K) for an extremely short length of 5 cm. These analyses were carried out using both the Supermagnet(trade mark, serif) and Venecia codes and the comparisons of the results are also discussed.« less

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

Jayakumar, R.; Martovetsky, N.N.; Perfect, S.A.

A glass-polyimide insulation system has been proposed by the US team for use in the Central Solenoid (CS) coil of the international Thermonuclear Experimental Reactor (ITER) machine and it is planned to use this system in the CS model coil inner module. The turn insulation will consist of 2 layers of combined prepreg and Kapton. Each layer is 50% overlapped with a butt wrap of prepreg and an overwrap of S glass. The coil layers will be separated by a glass-resin composite and impregnated in a VPI process. Small scale tests on the various components of the insulation are complete.more » It is planned to fabricate and test the insulation in a 4 x 4 insulated CS conductor array which will include the layer insulation and be vacuum impregnated. The conductor array will be subjected to 20 thermal cycles and 100000 mechanical load cycles in a Liquid Nitrogen environment. These loads are similar to those seen in the CS coil design. The insulation will be electrically tested at several stages during mechanical testing. This paper will describe the array configuration, fabrication: process, instrumentation, testing configuration, and supporting analyses used in selecting the array and test configurations.« less

Static magnetic fields: A summary of biological interactions, potential health effects, and exposure guidelines

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

Tenforde, T.S.

1992-05-01

Interest in the mechanisms of interaction and the biological effects of static magnetic fields has increased significantly during the past two decades as a result of the growing number of applications of these fields in research, industry and medicine. A major stimulus for research on the bioeffects of static magnetic fields has been the effort to develop new technologies for energy production and storage that utilize intense magnetic fields (e.g., thermonuclear fusion reactors and superconducting magnet energy storage devices). Interest in the possible biological interactions and health effects of static magnetic fields has also been increased as a result ofmore » recent developments in magnetic levitation as a mode of public transportation. In addition, the rapid emergence of magnetic resonance imaging as a new clinical diagnostic procedure has, in recent years, provided a strong rationale for defining the possible biological effects of magnetic fields with high flux densities. In this review, the principal interaction mechanisms of static magnetic fields will be described, and a summary will be given of the present state of knowledge of the biological, environmental, and human health effects of these fields.« less

Thermal analysis of the cryostat feed through for the ITER Tokamak TF feeder

NASA Astrophysics Data System (ADS)

Zhang, Shanwen; Song, Yuntao; Lu, Kun; Wang, Zhongwei; Zhang, Jianfeng; Qin, Yongfa

2017-04-01

In Tokamaks, the toroidal field (TF) coil feeder is an important component that is used to supply the cryogens and electrical power for the TF coils. As a part of the TF feeder, the cryostat-feed through (CFT) is subject to low temperatures of 9 and 80 K inside and room temperature of 300 K outside. Based on the features of the International Thermonuclear Experimental Reactor TF feeder, the thermal performance of the CFT under the nominal conditions is studied. Taking into account the conductive, convective and radiation heat transfer, the finite element model of the CFT is built. Transient thermal analysis is performed to determine the temperatures of the CFT on the 9th day of cooldown. The model is assessed by comparing the cooling curves of the CFT after 9 days. If the simulation and experimental results are the same, the finite element model can be considered as calibrated. The model predicts that the cooling time will be approximately 26 days and the temperature distribution and heat load of the main components are obtained when the CFT reaches thermal equilibrium. This study provides a valid quantitative characterization of the CFT design.

Pellet Injection in ITER with ∇B-induced Drift Effect using TASK/TR and HPI2 Codes

NASA Astrophysics Data System (ADS)

Kongkurd, R.; Wisitsorasak, A.

2017-09-01

The impact of pellet injection in International Thermonuclear Experimental Reactor (ITER) are investigated using integrated predictive modeling codes TASK/TR and HPI2 . In the core, the plasma profiles are predicted by the TASK/TR code in which the core transport models consist of a combination of the MMM95 anomalous transport model and NCLASS neoclassical transport. The pellet ablation in the plasma is described using neutral gas shielding (NGS) model with inclusion of the ∇B-induced \\overrightarrow{E}× \\overrightarrow{B} drift of the ionized ablated pellet particles. It is found that the high-field-side injection can deposit the pellet mass deeper than the injection from the low-field-side due to the advantage of the ∇B-induced drift. When pellets with deuterium-tritium mixing ratio of unity are launched with speed of 200 m/s, radius of 3 mm and injected at frequency of 2 Hz, the line average density and the plasma stored energy are increased by 80% and 25% respectively. The pellet material is mostly deposited at the normalized minor radius of 0.5 from the edge.

Evaluation of power transfer efficiency for a high power inductively coupled radio-frequency hydrogen ion source

NASA Astrophysics Data System (ADS)

Jain, P.; Recchia, M.; Cavenago, M.; Fantz, U.; Gaio, E.; Kraus, W.; Maistrello, A.; Veltri, P.

2018-04-01

Neutral beam injection (NBI) for plasma heating and current drive is necessary for International Thermonuclear Experimental reactor (ITER) tokamak. Due to its various advantages, a radio frequency (RF) driven plasma source type was selected as a reference ion source for the ITER heating NBI. The ITER relevant RF negative ion sources are inductively coupled (IC) devices whose operational working frequency has been chosen to be 1 MHz and are characterized by high RF power density (˜9.4 W cm-3) and low operational pressure (around 0.3 Pa). The RF field is produced by a coil in a cylindrical chamber leading to a plasma generation followed by its expansion inside the chamber. This paper recalls different concepts based on which a methodology is developed to evaluate the efficiency of the RF power transfer to hydrogen plasma. This efficiency is then analyzed as a function of the working frequency and in dependence of other operating source and plasma parameters. The study is applied to a high power IC RF hydrogen ion source which is similar to one simplified driver of the ELISE source (half the size of the ITER NBI source).

Off-design performance of a chemical looping combustion (CLC) combined cycle: effects of ambient temperature

NASA Astrophysics Data System (ADS)

Chi, Jinling; Wang, Bo; Zhang, Shijie; Xiao, Yunhan

2010-02-01

The present work investigates the influence of ambient temperature on the steady-state off-design thermodynamic performance of a chemical looping combustion (CLC) combined cycle. A sensitivity analysis of the CLC reactor system was conducted, which shows that the parameters that influence the temperatures of the CLC reactors most are the flow rate and temperature of air entering the air reactor. For the ambient temperature variation, three off-design control strategies have been assumed and compared: 1) without any Inlet Guide Vane (IGV) control, 2) IGV control to maintain air reactor temperature and 3) IGV control to maintain constant fuel reactor temperature, aside from fuel flow rate adjusting. Results indicate that, compared with the conventional combined cycle, due to the requirement of pressure balance at outlet of the two CLC reactors, CLC combined cycle shows completely different off-design thermodynamic characteristics regardless of the control strategy adopted. For the first control strategy, temperatures of the two CLC reactors both rise obviously as ambient temperature increases. IGV control adopted by the second and the third strategy has the effect to maintain one of the two reactors' temperatures at design condition when ambient temperature is above design point. Compare with the second strategy, the third would induce more severe decrease of efficiency and output power of the CLC combined cycle.

LIFE Materials: Overview of Fuels and Structural Materials Issues Volume 1

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

Farmer, J

2008-09-08

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spentmore » nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including un-enriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to divert in large quantities. Several topical reports are being prepared on the materials and processes required for the LIFE engine. Specific materials of interest include: (1) Baseline TRISO Fuel (TRISO); (2) Inert Matrix Fuel (IMF) & Other Alternative Solid Fuels; (3) Beryllium (Be) & Molten Lead Blankets (Pb/PbLi); (4) Molten Salt Coolants (FLIBE/FLiNaBe/FLiNaK); (5) Molten Salt Fuels (UF4 + FLIBE/FLiNaBe); (6) Cladding Materials for Fuel & Beryllium; (7) ODS FM Steel (ODS); (8) Solid First Wall (SFW); and (9) Solid-State Tritium Storage (Hydrides).« less

The secret of the Soviet hydrogen bomb

NASA Astrophysics Data System (ADS)

Wellerstein, Alex; Geist, Edward

2017-11-01

Was the first Soviet thermonuclear device really a step in the wrong direction? No bomb design has been as much maligned or otherwise disparaged as the first Soviet thermonuclear weapon. Detonated in August 1953, the bomb, officially tested under the name RDS-6s but usually known as Sloika or "layer cake" (the name Andrei Sakharov coined for it), was nothing to sneeze at. Shown in Figure 1 and able to be dropped from aircraft, it released the explosive equivalent, or yield, of almost half a megaton of TNT. The result was a blazing fireball with 20 times the power of the bomb that leveled Nagasaki, Japan.

Extended fusion yield integral using pathway idea in case of Shock-compressed heated plasma

NASA Astrophysics Data System (ADS)

Kumar, Dilip; Haubold, Hans

The extended non-resonant thermonuclear reaction rate probability integral obtained in Haubold and Kumar [Haubold, H.J. and Kumar, D.: 2008, Extension of thermonuclear functions through the pathway model including Maxwell-Boltzmann and Tsallis distributions, Astroparticle Physics, 29, 70-76] is used to evaluate the fusion energy by itegrating it over temperature. The closed form representation of the extended reaction rate integral via Meijer's G-function is expressed as a solution of a homogeneous differential equation. A physical model of Guderley[Guderley G. :1942, Starke kugelige und zylindrische Verdichtungsstsse in der Nhe des Kugelmittelpunktes bzw. der Zylinderachse, Luftfahrtforschung, 19, 302] has been considered for the laser driven hydrodynamical process in a compressed fusion plasma and heated strong spherical shock wave. The fusion yield integral obtained in the paper is compared with the standard fusion yield ob-tained by Haubold and John [Haubold, H.J. and John, R.W.:1981, Analytical representation of the thermonuclear reaction rate and fusion energy production in a spherical plasma shock wave, Plasma Physics, 5, 399-411]. The pathway parameter used in this paper is given an interpretation in terms of moments.

The classification of magnetohydrodynamic regimes of thermonuclear combustion

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

Remming, Ian S.; Khokhlov, Alexei M.

2014-10-10

Physical properties of magnetohydrodynamic (MHD) reaction fronts are studied as functions of the thermodynamic conditions, and the strength and orientation of the magnetic field in the unburned matter through which the fronts propagate. We determine the conditions for the existence of the various types of MHD reaction fronts and the character of the changes in physical quantities across these reaction fronts. The analysis is carried out in general for a perfect gas equation of state and a constant energy release, and then extended to thermonuclear reaction fronts in degenerate carbon-oxygen mixtures and degenerate helium in conditions typical of Type Iamore » supernova explosions. We find that as unburned matter enters perpendicular to a reaction front, the release of energy through burning generates shear velocity in the reacting gas that, depending on the type of reaction front, strengthens or weakens the magnetic field. In addition, we find that the steady-state propagation of a reaction front is impossible for certain ranges of magnetic field direction. Our results provide insight into the phenomena of MHD thermonuclear combustion that is relevant to the interpretation of future simulations of SN Ia explosions that have magnetic fields systematically incorporated.« less

Thermonuclear Explosions from Hybrid C/O/Ne White Dwarf Progenitors Ignited Centrally After Interior Mixing

NASA Astrophysics Data System (ADS)

Augustine, Carlyn

2018-01-01

Type Ia Supernovae are thermonuclear explosions of white dwarf (WD) stars. Past studies predict the existence of "hybrid" white dwarfs, made of a C/O/Ne core with a O/Ne shell, and that these are viable progenitors for supernovae. More recent work found that the C/O core is mixed with the surrounding O/Ne while the WD cools. Inspired by this scenario, we performed simulations of thermonuclear supernovae in the single degenerate paradigm from these hybrid progenitors. Our investigation began by constructing a hybrid white dwarf model with the one-dimensional stellar evolution code MESA. The model was allowed to go through unstable interior mixing ignite carbon burning centrally. The MESA model was then mapped to a two-dimensional initial condition and an explosion simulated from that with FLASH. For comparison, a similar simulation of an explosion was performed from a traditional C/O progenitor WD. Comparing the yields produced by explosion simulations allows us to determine which model produces more 56Ni, and therefore brighter events, and how explosions from these models differ from explosions from previous models without the mixing during the WD cooling.

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

PubMed

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

2005-07-01

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

Investigation on microstructure and properties of narrow-gap laser welding on reduced activation ferritic/martensitic steel CLF-1 with a thickness of 35 mm

NASA Astrophysics Data System (ADS)

Wu, Shikai; Zhang, Jianchao; Yang, Jiaoxi; Lu, Junxia; Liao, Hongbin; Wang, Xiaoyu

2018-05-01

Reduced activation ferritic martensitic (RAFM) steel is chosen as a structural material for test blanket modules (TBMs) to be constructed in International Thermonuclear Experimental Reactor (ITER) and China Fusion Engineering Test Reactor (CFETR). Chinese specific RAFM steel named with CLF-1 has been developed for CFETR. In this paper, a narrow-gap groove laser multi-pass welding of CLF-1 steel with thickness of 35 mm is conduced by YLS-15000 fiber laser. Further, the microstructures of different regions in the weld joint were characterized, and tensile impact and micro-hardness tests were carried out for evaluating the mecharical properties. The results show that the butt weld joint of CLF-1 steel with a thickness of 35 mm was well-formed using the optimal narrow-gap laser filler wire welding and no obvious defects was found such as incomplete fusion cracks and pores. The microstructures of backing layer is dominated by lath martensites and the Heat-Affected Zone (HAZ) was mainly filled with two-phase hybrid structures of secondary-tempering sorbites and martensites. The filler layer is similar to the backing layer in microstructures. In tensile tests, the tensile samples from different parts of the joint all fractured at base metal (BM). The micro-hardness of weld metal (WM) was found to be higher than that of BM and the Heat-Affected Zone (HAZ) exhibited no obvious softening. After post weld heat treatment (PWHT), it can be observed that the fusion zone of the autogenous welding bead and the upper filling beads mainly consist of lath martensites which caused the lower impact absorbing energy. The HAZ mainly included two-phase hybrid structures of secondary-tempering sorbites and martensites and exhibited favorable impact toughness.

Magnet Design Considerations for Fusion Nuclear Science Facility

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

Zhai, Y.; Kessel, C.; El-Guebaly, L.

2016-06-01

The Fusion Nuclear Science Facility (FNSF) is a nuclear confinement facility that provides a fusion environment with components of the reactor integrated together to bridge the technical gaps of burning plasma and nuclear science between the International Thermonuclear Experimental Reactor (ITER) and the demonstration power plant (DEMO). Compared with ITER, the FNSF is smaller in size but generates much higher magnetic field, i.e., 30 times higher neutron fluence with three orders of magnitude longer plasma operation at higher operating temperatures for structures surrounding the plasma. Input parameters to the magnet design from system code analysis include magnetic field of 7.5more » T at the plasma center with a plasma major radius of 4.8 m and a minor radius of 1.2 m and a peak field of 15.5 T on the toroidal field (TF) coils for the FNSF. Both low-temperature superconductors (LTS) and high-temperature superconductors (HTS) are considered for the FNSF magnet design based on the state-of-the-art fusion magnet technology. The higher magnetic field can be achieved by using the high-performance ternary restacked-rod process Nb3Sn strands for TF magnets. The circular cable-in-conduit conductor (CICC) design similar to ITER magnets and a high-aspect-ratio rectangular CICC design are evaluated for FNSF magnets, but low-activation-jacket materials may need to be selected. The conductor design concept and TF coil winding pack composition and dimension based on the horizontal maintenance schemes are discussed. Neutron radiation limits for the LTS and HTS superconductors and electrical insulation materials are also reviewed based on the available materials previously tested. The material radiation limits for FNSF magnets are defined as part of the conceptual design studies for FNSF magnets.« less

Transportation properties of a high-current magnetically insulated transmission line and dynamics of the electrode plasma

NASA Astrophysics Data System (ADS)

Anan'ev, S. S.; Bakshaev, Yu. L.; Bartov, A. V.; Blinov, P. I.; Dan'ko, S. A.; Zhuzhunashvili, A. I.; Kazakov, E. D.; Kalinin, Yu. G.; Kingsep, A. S.; Korolev, V. D.; Mizhiritskii, V. I.; Smirnov, V. P.; Tkachenko, S. I.; Chernenko, A. S.

2008-07-01

Results are presented from experimental studies of a section of a magnetically insulated transmission line (MITL) with a current density of up to 500 MA/cm2 and linear current density of up to 7 MA/cm (the parameters close to those in a fast-Z-pinch-driven fusion reactor projected at Sandia Laboratories). The experiments were performed in the S-300 facility (3 MA, 0.15 Ω, 100 ns). At high linear current densities, the surface of the ohmically heated MITL electrode can explode and a plasma layer can form near the electrode surface. As a result, the MITL can lose its transmission properties due to the shunting of the vacuum gap by the plasma produced. In this series of experiments, the dynamics of the electrode plasma and the dependence of the transmission properties of the MITL on the material and cleanness of the electrode surface were studied. It is shown experimentally that, when the current with a linear density of up to 7 MA/cm begins to flow along a model MITL, the input and output currents differ by less than 10% over a time interval of up to 230 ns for nickel electrodes and up to 350 ns for a line with a gold central electrode. No effect of the oil film present on the electrode surface on the loss of the transmission properties of the line was observed. It is also shown that electron losses insignificantly contribute to the total current balance. The experimental results are compared with calculations of the electrode explosion and the subsequent expansion of the plasma layer. A conclusion is made that the life-time of the model MITL satisfies the requirements imposed on the transmission lines intended for use in the projected thermonuclear reactor.

Modeling and simulation of CANDU reactor and its regulating system

NASA Astrophysics Data System (ADS)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different phenomena related to the transfer of the energy from the core. The main function of the reactor regulating system is to control the power of the reactor. This is achieved by using a set of detectors. reactivity devices. and digital control algorithms. Three main reactivity devices that are activated during short-term or intermediate-term transients are modeled in this thesis. The main elements of the digital control system are implemented in accordance to the program specifications for the actual control system in CANDU reactors. The simulation results are validated against requirements of the reactor regulating system. actual plant data. and pre-validated data from other computer codes. The validation process shows that the simulation results can be trusted in making engineering decisions regarding the reactor regulating system and prediction of the system performance in response to upset conditions or disturbances. KEYWORDS: CANDU reactors. reactor regulating system. nodal model. spatial kinetics. reactivity devices. simulation.

NEUTRONIC REACTOR SYSTEM

DOEpatents

Treshow, M.

1959-02-10

A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

Enhanced biodegradation of hexachlorocyclohexane in upflow anaerobic sludge blanket reactor using methanol as an electron donor.

PubMed

Bhatt, Praveena; Kumar, M Suresh; Mudliar, Sandeep; Chakrabarti, Tapan

2008-05-01

Anaerobic dechlorination of technical grade hexachlorocyclohexane (THCH) was studied in a continuous upflow anaerobic sludge blanket (UASB) reactor with methanol as a supplementary substrate and electron donor. A reactor without methanol served as the experimental control. The inlet feed concentration of THCH in both the experimental and the control UASB reactor was 100 mg l(-1). After 60 days of continuous operation, the removal of THCH was >99% in the methanol-supplemented reactor as compared to 20-35% in the control reactor. THCH was completely dechlorinated in the methanol fed reactor at 48 h HRT after 2 months of continuous operation. This period was also accompanied by increase in biomass in the reactor, which was not observed in the experimental control. Batch studies using other supplementary substrates as well as electron donors namely acetate, butyrate, formate and ethanol showed lower % dechlorination (<85%) and dechlorination rates (<3 mg g(-1)d(-1)) as compared to methanol (98%, 5 mg g(-1)d(-1)). The optimum concentration of methanol required, for stable dechlorination of THCH (100 mg l(-1)) in the UASB reactor, was found to be 500 mg l(-1). Results indicate that addition of methanol as electron donor enhances dechlorination of THCH at high inlet concentration, and is also required for stable UASB reactor performance.

Effects of imperfect mixing on low-density polyethylene reactor dynamics

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

Villa, C.M.; Dihora, J.O.; Ray, W.H.

1998-07-01

Earlier work considered the effect of feed conditions and controller configuration on the runaway behavior of LDPE autoclave reactors assuming a perfectly mixed reactor. This study provides additional insight on the dynamics of such reactors by using an imperfectly mixed reactor model and bifurcation analysis to show the changes in the stability region when there is imperfect macroscale mixing. The presence of imperfect mixing substantially increases the range of stable operation of the reactor and makes the process much easier to control than for a perfectly mixed reactor. The results of model analysis and simulations are used to identify somemore » of the conditions that lead to unstable reactor behavior and to suggest ways to avoid reactor runaway or reactor extinction during grade transitions and other process operation disturbances.« less

Control rod system useable for fuel handling in a gas-cooled nuclear reactor

DOEpatents

Spurrier, Francis R.

1976-11-30

A control rod and its associated drive are used to elevate a complete stack of fuel blocks to a position above the core of a gas-cooled nuclear reactor. A fuel-handling machine grasps the control rod and the drive is unlatched from the rod. The stack and rod are transferred out of the reactor, or to a new location in the reactor, by the fuel-handling machine.

Dual annular rotating [open quotes]windowed[close quotes] nuclear reflector reactor control system

DOEpatents

Jacox, M.G.; Drexler, R.L.; Hunt, R.N.M.; Lake, J.A.

1994-03-29

A nuclear reactor control system is provided in a nuclear reactor having a core operating in the fast neutron energy spectrum where criticality control is achieved by neutron leakage. The control system includes dual annular, rotatable reflector rings. There are two reflector rings: an inner reflector ring and an outer reflector ring. The reflectors are concentrically assembled, surround the reactor core, and each reflector ring includes a plurality of openings. The openings in each ring are capable of being aligned or non-aligned with each other. Independent driving means for each of the annular reflector rings is provided so that reactor criticality can be initiated and controlled by rotation of either reflector ring such that the extent of alignment of the openings in each ring controls the reflection of neutrons from the core. 4 figures.

156. ARAIII Reactor building (ARA608) Electrical and control details of ...

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

156. ARA-III Reactor building (ARA-608) Electrical and control details of mobile work bridge over reactor and pipiing pits. Aerojet-general 880-area/GCRE-608-E-6. Date: November 1958. Ineel index code no. 063-0608-10-013-102621. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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

NONE

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

Adaptive control method for core power control in TRIGA Mark II reactor

NASA Astrophysics Data System (ADS)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

77 FR 58419 - Guidelines for Preparing and Reviewing Licensing Applications for Instrumentation and Control...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-20

... Applications for Instrumentation and Control Upgrades for Non-Power Reactors AGENCY: Nuclear Regulatory...-Power Reactors: Format and Content,'' for instrumentation and control upgrades and NUREG-1537, Part 2, ``Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors: Standard Review...

Integrated head package for top mounted nuclear instrumentation

DOEpatents

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

1993-01-01

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

Control system for a small fission reactor

DOEpatents

Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

1985-02-08

A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

Reactor transient control in support of PFR/TREAT TUCOP experiments

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

Burrows, D.R.; Larsen, G.R.; Harrison, L.J.

1984-01-01

Unique energy deposition and experiment control requirements posed bythe PFR/TREAT series of transient undercooling/overpower (TUCOP) experiments resulted in equally unique TREAT reactor operations. New reactor control computer algorithms were written and used with the TREAT reactor control computer system to perform such functions as early power burst generation (based on test train flow conditions), burst generation produced by a step insertion of reactivity following a controlled power ramp, and shutdown (SCRAM) initiators based on both test train conditions and energy deposition. Specialized hardware was constructed to simulate test train inputs to the control computer system so that computer algorithms couldmore » be tested in real time without irradiating the experiment.« less

Overview of Indian activities on fusion reactor materials

NASA Astrophysics Data System (ADS)

Banerjee, Srikumar

2014-12-01

This paper on overview of Indian activities on fusion reactor materials describes in brief the efforts India has made to develop materials for the first wall of a tokamak, its blanket and superconducting magnet coils. Through a systematic and scientific approach, India has developed and commercially produced reduced activation ferritic/martensitic (RAFM) steel that is comparable to Eurofer 97. Powder of low activation ferritic/martensitic oxide dispersion strengthened steel with characteristics desired for its application in the first wall of a tokamak has been produced on the laboratory scale. V-4Cr-4Ti alloy was also prepared in the laboratory, and kinetics of hydrogen absorption in this was investigated. Cu-1 wt%Cr-0.1 wt%Zr - an alloy meant for use as heat transfer elements for hypervapotrons and heat sink for the first wall - was developed and characterized in detail for its aging behavior. The role of addition of a small quantity of Zr in its improved fatigue performance was delineated, and its diffusion bonding with both W and stainless steel was achieved using Ni as an interlayer. The alloy was produced in large quantities and used for manufacturing both the heat transfer elements and components for the International Thermonuclear Experimental Reactor (ITER). India has proposed to install and test a lead-lithium cooled ceramic breeder test blanket module (LLCB-TBM) at ITER. To meet this objective, efforts have been made to produce and characterize Li2TiO3 pebbles, and also improve the thermal conductivity of packed beds of these pebbles. Liquid metal loops have been set up and corrosion behavior of RAFM steel in flowing Pb-Li eutectic has been studied in the presence as well as absence of magnetic fields. To prevent permeation of tritium and reduce the magneto-hydro-dynamic drag, processes have been developed for coating alumina on RAFM steel. Apart from these activities, different approaches being attempted to make the U-shaped first wall of the TBM box are briefly described. India has also initiated the development of fusion grade superconductors. Success achieved in the fabrication of Nb3Sn based multi-filamentary wires using the internal tin process and cable-in-conduit-conductors is also briefly presented.

Closed fuel cycle with increased fuel burn-up and economy applying of thorium resources

NASA Astrophysics Data System (ADS)

Kulikov, G. G.; Apse, V. A.

2017-01-01

The possible role of existing thorium reserves in the Russian Federation on engaging thorium in being currently closed (U-Pu)-fuel cycle of nuclear power of the country is considered. The application efficiency of thermonuclear neutron sources with thorium blanket for the economical use of existing thorium reserves is demonstrated. The aim of the work is to find solutions of such major tasks as the reduction of both front-end and back-end of nuclear fuel cycle and an enhancing its protection against the uncontrolled proliferation of fissile materials by means of the smallest changes in the fuel cycle. During implementation of the work we analyzed the results obtained earlier by the authors, brought new information on the number of thorium available in the Russian Federation and made further assessments. On the basis of proposal on the inclusion of hybrid reactors with Th-blanket into the future nuclear power for the production of light uranium fraction 232+233+234U, and 231Pa, we obtained the following results: 1. The fuel cycle will shift from fissile 235U to 233U which is more attractive for thermal power reactors. 2. The light uranium fraction is the most "protected" in the uranium component of fuel and mixed with regenerated uranium will in addition become a low enriched uranium fuel, that will weaken the problem of uncontrolled proliferation of fissile materials. 3. 231Pa doping into the fuel stabilizes its multiplying properties that will allow us to implement long-term fuel residence time and eventually to increase the export potential of all nuclear power technologies. 4. The thorium reserves being near city Krasnoufimsk (Russia) are large enough for operation of large-scale nuclear power of the Russian Federation of 70 GWe capacity during more than a quarter century under assumption that thorium is loaded into blankets of hybrid TNS only. The general conclusion: the inclusion of a small number of hybrid reactors with Th-blanket into the future nuclear power will allow us substantially to solve its problems, as well as to increase its export potential.

Autonomous Control of Nuclear Power Plants

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

Basher, H.

2003-10-20

A nuclear reactor is a complex system that requires highly sophisticated controllers to ensure that desired performance and safety can be achieved and maintained during its operations. Higher-demanding operational requirements such as reliability, lower environmental impacts, and improved performance under adverse conditions in nuclear power plants, coupled with the complexity and uncertainty of the models, necessitate the use of an increased level of autonomy in the control methods. In the opinion of many researchers, the tasks involved during nuclear reactor design and operation (e.g., design optimization, transient diagnosis, and core reload optimization) involve important human cognition and decisions that maymore » be more easily achieved with intelligent methods such as expert systems, fuzzy logic, neural networks, and genetic algorithms. Many experts in the field of control systems share the idea that a higher degree of autonomy in control of complex systems such as nuclear plants is more easily achievable through the integration of conventional control systems and the intelligent components. Researchers have investigated the feasibility of the integration of fuzzy logic, neural networks, genetic algorithms, and expert systems with the conventional control methods to achieve higher degrees of autonomy in different aspects of reactor operations such as reactor startup, shutdown in emergency situations, fault detection and diagnosis, nuclear reactor alarm processing and diagnosis, and reactor load-following operations, to name a few. With the advancement of new technologies and computing power, it is feasible to automate most of the nuclear reactor control and operation, which will result in increased safety and economical benefits. This study surveys current status, practices, and recent advances made towards developing autonomous control systems for nuclear reactors.« less

REACTOR CONTROL SYSTEM

DOEpatents

MacNeill, J.H.; Estabrook, J.Y.

1960-05-10

A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

NEUTRONIC REACTOR CONTROL

DOEpatents

Metcalf, H.E.

1958-10-14

Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

Inhibition of nitrification in municipal wastewater-treating photobioreactors: Effect on algal growth and nutrient uptake.

PubMed

Krustok, I; Odlare, M; Truu, J; Nehrenheim, E

2016-02-01

The effect of inhibiting nitrification on algal growth and nutrient uptake was studied in photobioreactors treating municipal wastewater. As previous studies have indicated that algae prefer certain nitrogen species to others, and because nitrifying bacteria are inhibited by microalgae, it is important to shed more light on these interactions. In this study allylthiourea (ATU) was used to inhibit nitrification in wastewater-treating photobioreactors. The nitrification-inhibited reactors were compared to control reactors with no ATU added. Microalgae had higher growth in the inhibited reactors, resulting in a higher chlorophyll a concentration. The species mix also differed, with Chlorella and Scenedesmus being the dominant genera in the control reactors and Cryptomonas and Chlorella dominating in the inhibited reactors. The nitrogen speciation in the reactors after 8 days incubation was also different in the two setups, with N existing mostly as NH4-N in the inhibited reactors and as NO3-N in the control reactors. Copyright © 2015 Elsevier Ltd. All rights reserved.

NUCLEAR REACTOR

DOEpatents

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Achieving nitritation in a continuous moving bed biofilm reactor at different temperatures through ratio control.

PubMed

Bian, Wei; Zhang, Shuyan; Zhang, Yanzhuo; Li, Wenjing; Kan, Ruizhe; Wang, Wenxiao; Zheng, Zhaoming; Li, Jun

2017-02-01

A ratio control strategy was implemented in a continuous moving bed biofilm reactor (MBBR) to investigate the response to different temperatures. The control strategy was designed to maintain a constant ratio between dissolved oxygen (DO) and total ammonia nitrogen (TAN) concentrations. The results revealed that a stable nitritation in a biofilm reactor could be achieved via ratio control, which compensated the negative influence of low temperatures by stronger oxygen-limiting conditions. Even with a temperature as low as 6°C, stable nitritation could be achieved when the controlling ratio did not exceed 0.17. Oxygen-limiting conditions in the biofilm reactor were determined by the DO/TAN concentrations ratio, instead of the mere DO concentration. This ratio control strategy allowed the achievement of stable nitritation without complete wash-out of NOB from the reactor. Through the ratio control strategy full nitritation of sidestream wastewater was allowed; however, for mainstream wastewater, only partial nitritation was recommended. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

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

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

2009-09-15

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

Hydraulic Actuator for Ganged Control Rods

NASA Technical Reports Server (NTRS)

Thompson, D. C.; Robey, R. M.

1986-01-01

Hydraulic actuator moves several nuclear-reactor control rods in unison. Electromagnetic pump pushes liquid lithium against ends of control rods, forcing them out of or into nuclear reactor. Color arrows show lithium flow for reactor startup and operation. Flow reversed for shutdown. Conceived for use aboard spacecraft, actuator principle applied to terrestrial hydraulic machinery involving motion of ganged rods.

CONTROL SYSTEM

DOEpatents

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

1962-10-30

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

Reactor control rod timing system. [LMFBR

DOEpatents

Wu, P.T.K.

1980-03-18

A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

Reactor control rod timing system

DOEpatents

Wu, Peter T. K.

1982-01-01

A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

ETR CONTROL BUILDING, TRA647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT ...

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

ETR CONTROL BUILDING, TRA-647, INTERIOR. CONTROL ROOM, CONTEXTUAL VIEW. INSTRUMENT PANELS AT REAR OF OPERATOR'S CONSOLE GAVE OPERATOR STATUS OF REACTOR PERFORMANCE, COOLANT-WATER CHARACTERISTICS AND OTHER INDICATORS. WINDOWS AT RIGHT LOOKED INTO ETR BUILDING FIRST FLOOR. CAMERA FACING EAST. INL NEGATIVE NO. HD42-6. Mike Crane, Photographer, 3/2004 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Nuclear electric propulsion reactor control systems status

NASA Technical Reports Server (NTRS)

Ferg, D. A.

1973-01-01

The thermionic reactor control system design studies conducted over the past several years for a nuclear electric propulsion system are described and summarized. The relevant reactor control system studies are discussed in qualitative terms, pointing out the significant advantages and disadvantages including the impact that the various control systems would have on the nuclear electric propulsion system design. A recommendation for the reference control system is made, and a program for future work leading to an engineering model is described.

REACTOR CONTROL

DOEpatents

Fortescue, P.; Nicoll, D.

1962-04-24

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Quick release latch for reactor scram

DOEpatents

Johnson, Melvin L.; Shawver, Bruce M.

1976-01-01

A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet-type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel.

Quick release latch for reactor scram

DOEpatents

Johnson, M.L.; Shawver, B.M.

1975-09-16

A simple, reliable, and fast-acting means for releasing a control element and allowing it to be inserted rapidly into the core region of a nuclear reactor for scram purposes is described. A latch mechanism grips a coupling head on a nuclear control element to connect the control element to the control drive assembly. The latch mechanism is closed by tensioning a cable or rod with an actuator. The control element is released by de-energizing the actuator, providing fail-safe, rapid release of the control element to effect reactor shutdown. A sensing rod provides indication that the control element is properly positioned in the latch. Two embodiments are illustrated, one involving a collet- type latch mechanism, the other a pliers-type latch mechanism with the actuator located inside the reactor vessel. (auth)

Effects of Thermonuclear X-Ray Bursts on Non-burst Emissions in the Soft State of 4U 1728–34

NASA Astrophysics Data System (ADS)

Bhattacharyya, Sudip; Yadav, J. S.; Sridhar, Navin; Verdhan Chauhan, Jai; Agrawal, P. C.; Antia, H. M.; Pahari, Mayukh; Misra, Ranjeev; Katoch, Tilak; Manchanda, R. K.; Paul, Biswajit

2018-06-01

It has recently been shown that the persistent emission of a neutron star low-mass X-ray binary (LMXB) evolves during a thermonuclear (type-I) X-ray burst. The reason of this evolution, however, is not fully known. This uncertainty can introduce significant systematics in the neutron star radius measurement using burst spectra, particularly if an unknown but significant fraction of the burst emission, which is reprocessed, contributes to the changes in the persistent emission during the burst. Here, by analyzing individual burst data of AstroSat/LAXPC from the neutron star LMXB 4U 1728–34 in the soft state, we show that the burst emission is not significantly reprocessed by a corona covering the neutron star. Rather, our analysis suggests that the burst emission enhances the accretion disk emission, possibly by increasing the accretion rate via disk. This enhanced disk emission, which is Comptonized by a corona covering the disk, can explain an increased persistent emission observed during the burst. This finding provides an understanding of persistent emission components and their interaction with the thermonuclear burst emission. Furthermore, as burst photons are not significantly reprocessed, non-burst and burst emissions can be reliably separated, which is required to reduce systematic uncertainties in the stellar radius measurement.

Oscillations During Thermonuclear X-ray Bursts

NASA Technical Reports Server (NTRS)

Strohmayer, Tod E.; White, Nicholas E. (Technical Monitor)

2001-01-01

High amplitude, nearly coherent X-ray brightness oscillations during thermonuclear X-ray bursts were discovered with the Rossi X-ray Timing Explorer (RXTE) in early 1996. Spectral and timing evidence strongly supports the conclusion that these oscillations are caused by rotational modulation of the burst emission and that they reveal the spin frequency of neutron stars in low mass X-ray binaries, a long sought goal of X-ray astronomy. Studies carried out over the past year have led to the discovery of burst oscillations in four new sources, bringing to ten the number with confirmed burst oscillations. I review the status of our knowledge of these oscillations and indicate how they can be used to probe the physics of neutron stars. For a few burst oscillation sources it has been proposed that the strongest and most ubiquitous frequency is actually the first overtone of the spin frequency and hence that two nearly antipodal hot spots are present on the neutron star. This inference has important implications for both the physics of thermonuclear burning as well as the mass - radius relation for neutron stars, so its confirmation is crucial. I discuss recent attempts to confirm this hypothesis for 4U 1636-53, the source for which a signal at the putative fundamental (290Hz) has, been claimed.

Search for stellar collapse with the MACRO detector at Gran Sasso

NASA Technical Reports Server (NTRS)

Steinberg, R.

1985-01-01

It is viewed that in stellar evolution stars in the range of 8 to 12 solar masses evolve gradually as increasingly heavier nuclei are produced and then consumed in a series of exothermic thermonuclear processes ultimately leading to the formation of a core composed almost entirely of nickel and iron. When the mass of this hot iron-nickel core reaches the critical value of approximately 1.4 solar masses, electron degeneracy pressure is no longer able to support the outer layers of the star and a collapse process begins. Since the core has exhausted its thermonuclear fuel, further stages of thermonuclear burning cannot prevent a runaway collapse. As the density reaches 10 to the 10th power gm sub/cm at a temperature near 10 to the 10th power k, most of the heavy nuclei are dissociated into free nucleons and electron capture on free protons leads to a decrease in the degeneracy pressure and further acceleration of the collapse process. Although this general picture has received substantial confirmation over the past two decades with the discovery of radio pulsars (neutron stars), X-ray pulsars (accreting binary neutron stars) and Cyg X-1 (probably an accreting black hole), an actual neutrino burst is not yet convincingly detected.

Vulnerability assessment of a space based weapon platform electronic system exposed to a thermonuclear weapon detonation

NASA Astrophysics Data System (ADS)

Perez, C. L.; Johnson, J. O.

Rapidly changing world events, the increased number of nations with inter-continental ballistic missile capability, and the proliferation of nuclear weapon technology will increase the number of nuclear threats facing the world today. Monitoring these nation's activities and providing an early warning and/or intercept system via reconnaissance and surveillance satellites and space based weapon platforms is a viable deterrent against a surprise nuclear attack. However, the deployment of satellite and weapon platform assets in space will subject the sensitive electronic equipment to a variety of natural and man-made radiation environments. These include Van Allen Belt protons and electrons; galactic and solar flare protons; and neutrons, gamma rays, and x-rays from intentionally detonated fission and fusion weapons. In this paper, the MASH vl.0 code system is used to estimate the dose to the critical electronics components of an idealized space based weapon platform from neutron and gamma-ray radiation emitted from a thermonuclear weapon detonation in space. Fluence and dose assessments were performed for the platform fully loaded, and in several stages representing limited engagement scenarios. The results indicate vulnerabilities to the Command, Control, and Communication bay instruments from radiation damage for a nuclear weapon detonation for certain source/platform orientations. The distance at which damage occurs will depend on the weapon yield (n,(gamma)/kiloton) and size (kilotons).

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

USGS Publications Warehouse

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

2016-01-01

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

An adaptive load-following control system for a space nuclear power system

NASA Astrophysics Data System (ADS)

Metzger, John D.; El-Genk, Mohamed S.

An adaptive load-following control system is proposed for a space nuclear power system. The conceptual design of the SP-100 space nuclear power system proposes operating the nuclear reactor at a base thermal power and accommodating changes in the electrical power demand with a shunt regulator. It is necessary to increase the reactor thermal power if the payload electrical demand exceeds the peak system electrical output for the associated reactor power. When it is necessary to change the nuclear reactor power to meet a change in the power demand, the power ascension or descension must be accomplished in a predetermined manner to avoid thermal stresses in the system and to achieve the desired reactor period. The load-following control system described has the ability to adapt to changes in the system and to changes in the satellite environment. The application is proposed of the model reference adaptive control (MRAC). The adaptive control system has the ability to control the dynamic response of nonlinear systems. Three basic subsets of adaptive control are: (1) gain scheduling, (2) self-tuning regulators, and (3) model reference adaptive control.

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

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

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

2014-11-11

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

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

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

Bernard, J.A.

1989-09-01

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

Pulse-density modulation control of chemical oscillation far from equilibrium in a droplet open-reactor system

PubMed Central

Sugiura, Haruka; Ito, Manami; Okuaki, Tomoya; Mori, Yoshihito; Kitahata, Hiroyuki; Takinoue, Masahiro

2016-01-01

The design, construction and control of artificial self-organized systems modelled on dynamical behaviours of living systems are important issues in biologically inspired engineering. Such systems are usually based on complex reaction dynamics far from equilibrium; therefore, the control of non-equilibrium conditions is required. Here we report a droplet open-reactor system, based on droplet fusion and fission, that achieves dynamical control over chemical fluxes into/out of the reactor for chemical reactions far from equilibrium. We mathematically reveal that the control mechanism is formulated as pulse-density modulation control of the fusion–fission timing. We produce the droplet open-reactor system using microfluidic technologies and then perform external control and autonomous feedback control over autocatalytic chemical oscillation reactions far from equilibrium. We believe that this system will be valuable for the dynamical control over self-organized phenomena far from equilibrium in chemical and biomedical studies. PMID:26786848

Pulse-density modulation control of chemical oscillation far from equilibrium in a droplet open-reactor system.

PubMed

Sugiura, Haruka; Ito, Manami; Okuaki, Tomoya; Mori, Yoshihito; Kitahata, Hiroyuki; Takinoue, Masahiro

2016-01-20

The design, construction and control of artificial self-organized systems modelled on dynamical behaviours of living systems are important issues in biologically inspired engineering. Such systems are usually based on complex reaction dynamics far from equilibrium; therefore, the control of non-equilibrium conditions is required. Here we report a droplet open-reactor system, based on droplet fusion and fission, that achieves dynamical control over chemical fluxes into/out of the reactor for chemical reactions far from equilibrium. We mathematically reveal that the control mechanism is formulated as pulse-density modulation control of the fusion-fission timing. We produce the droplet open-reactor system using microfluidic technologies and then perform external control and autonomous feedback control over autocatalytic chemical oscillation reactions far from equilibrium. We believe that this system will be valuable for the dynamical control over self-organized phenomena far from equilibrium in chemical and biomedical studies.

Fluidic self-actuating control assembly

DOEpatents

Grantz, Alan L.

1979-01-01

A fluidic self-actuating control assembly for use in a reactor wherein no external control inputs are required to actuate (scram) the system. The assembly is constructed to scram upon sensing either a sudden depressurization of reactor inlet flow or a sudden increase in core neutron flux. A fluidic control system senses abnormal flow or neutron flux transients and actuates the system, whereupon assembly coolant flow reverses, forcing absorber balls into the reactor core region.

III International Conference on Laser and Plasma Researches and Technologies

NASA Astrophysics Data System (ADS)

2017-12-01

A.P. Kuznetsov and S.V. Genisaretskaya III Conference on Plasma and Laser Research and Technologies took place on January 24th until January 27th, 2017 at the National Research Nuclear University "MEPhI" (NRNU MEPhI). The Conference was organized by the Institute for Laser and Plasma Technologies and was supported by the Competitiveness Program of NRNU MEPhI. The conference program consisted of nine sections: • Laser physics and its application • Plasma physics and its application • Laser, plasma and radiation technologies in industry • Physics of extreme light fields • Controlled thermonuclear fusion • Modern problems of theoretical physics • Challenges in physics of solid state, functional materials and nanosystems • Particle accelerators and radiation technologies • Modern trends of quantum metrology. The conference is based on scientific fields as follows: • Laser, plasma and radiation technologies in industry, energetic, medicine; • Photonics, quantum metrology, optical information processing; • New functional materials, metamaterials, “smart” alloys and quantum systems; • Ultrahigh optical fields, high-power lasers, Mega Science facilities; • High-temperature plasma physics, environmentally-friendly energetic based on controlled thermonuclear fusion; • Spectroscopic synchrotron, neutron, laser research methods, quantum mechanical calculation and computer modelling of condensed media and nanostructures. More than 250 specialists took part in the Conference. They represented leading Russian scientific research centers and universities (National Research Centre "Kurchatov Institute", A.M. Prokhorov General Physics Institute, P.N. Lebedev Physical Institute, Troitsk Institute for Innovation and Fusion Research, Joint Institute for Nuclear Research, Moscow Institute of Physics and Tecnology and others) and leading scientific centers and universities from Germany, France, USA, Canada, Japan. We would like to thank heartily all of the speakers, participants, organizing and program committee members for their contribution to the conference.

METHOD FOR SENSING DEGREE OF FLUIDIZATION IN FLUIDIZED BED

DOEpatents

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

1961-12-12

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

10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

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

Bohachek, Randolph Charles

2015-09-01

The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactorsmore » is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.« less

NEUTRONIC REACTOR

DOEpatents

Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

1959-03-24

A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

Plum Brook Reactor Facility Control Room during Facility Startup

NASA Image and Video Library

1961-02-21

Operators test the National Aeronautics and Space Administration’s (NASA) Plum Brook Reactor Facility systems in the months leading up to its actual operation. The “Reactor On” signs are illuminated but the reactor core was not yet ready for chain reactions. Just a couple weeks after this photograph, Plum Brook Station held a media open house to unveil the 60-megawatt test reactor near Sandusky, Ohio. More than 60 members of the print media and radio and television news services met at the site to talk with community leaders and representatives from NASA and Atomic Energy Commission. The Plum Brook reactor went critical for the first time on the evening of June 14, 1961. It was not until April 1963 that the reactor reached its full potential of 60 megawatts. The reactor control room, located on the second floor of the facility, was run by licensed operators. The operators manually operated the shim rods which adjusted the chain reaction in the reactor core. The regulating rods could partially or completely shut down the reactor. The control room also housed remote area monitoring panels and other monitoring equipment that allowed operators to monitor radiation sensors located throughout the facility and to scram the reactor instantly if necessary. The color of the indicator lights corresponded with the elevation of the detectors in the various buildings. The reactor could also shut itself down automatically if the monitors detected any sudden irregularities.

Computer simulation of magnetization-controlled shunt reactors for calculating electromagnetic transients in power systems

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

Karpov, A. S.

2013-01-15

A computer procedure for simulating magnetization-controlled dc shunt reactors is described, which enables the electromagnetic transients in electric power systems to be calculated. It is shown that, by taking technically simple measures in the control system, one can obtain high-speed reactors sufficient for many purposes, and dispense with the use of high-power devices for compensating higher harmonic components.

Evidence for a link between atmospheric thermonuclear detonations and nitric acid.

PubMed

Holdsworth, G

1986-12-11

Suitably located glacier cores, obtained from high-altitude, low-temperature sites, can reveal detailed information about atmospheric air chemistry at sub-annual resolution 1 . Such data may provide input to climate-change models, the study of acid precipitation patterns and many other phenomena. Here I present data from an ice core which show that during the era of intense atmospheric thermonuclear weapons testing (ATWT) a significant part of the nitrate content in the snow was modulated by the intensity of the nuclear detonations. The fixation of nitrogen by nuclear fireballs leads to NO x gases in the atmosphere 2 and ultimately to nitric acid in precipitation. At certain concentrations, these gases and the associated aerosols may perturb the climate 3,4 .

NACRE II: an update of the NACRE compilation of charged-particle-induced thermonuclear reaction rates for nuclei with mass number A<16

NASA Astrophysics Data System (ADS)

Xu, Y.; Takahashi, K.; Goriely, S.; Arnould, M.; Ohta, M.; Utsunomiya, H.

2013-11-01

An update of the NACRE compilation [3] is presented. This new compilation, referred to as NACRE II, reports thermonuclear reaction rates for 34 charged-particle induced, two-body exoergic reactions on nuclides with mass number A<16, of which fifteen are particle-transfer reactions and the rest radiative capture reactions. When compared with NACRE, NACRE II features in particular (1) the addition to the experimental data collected in NACRE of those reported later, preferentially in the major journals of the field by early 2013, and (2) the adoption of potential models as the primary tool for extrapolation to very low energies of astrophysical S-factors, with a systematic evaluation of uncertainties.

A NON-PRE DOUBLE-PEAKED BURST FROM 4U 1636-536: EVIDENCE FOR BURNING FRONT PROPAGATION

NASA Technical Reports Server (NTRS)

Bhattacharyya, Sudip; Strohmayer, Tod E.

2005-01-01

We analyse Rossi X-ray Timing Explorer (RXTE) Proportional Counter Array (PCA) data of a double-peaked burst from the low mass X-ray binary (LMXB) 4U 1636-536 that shows no evidence for photospheric radius expansion (PRE). We find that the X-ray emitting area on the star increases with time as the burst progresses, even though the photosphere does not expand. We argue that this is a strong indication of thermonuclear flame spreading on the stellar surface during such bursts. We propose a model for such double-peaked bursts, based on thermonuclear flame spreading, that can qualitatively explain their essential features, as well as the rarity of these bursts.

Magnetic Reconnection Driven by Thermonuclear Burning

NASA Astrophysics Data System (ADS)

Gatto, R.; Coppi, B.

2017-10-01

Considering that fusion reaction products (e.g. α-particles) deposit their energy on the electrons, the relevant thermal energy balance equation is characterized by a fusion source term, a relatively large longitudinal thermal conductivity and an appropriate transverse thermal conductivity. Then, looking for modes that are radially localized around rational surfaces, reconnected field configurations are found that can be sustained by the electron thermal energy source due to fusion reactions. Then this process can be included in the category of endogenous reconnection processes and may be viewed as a form of the thermonuclear instability that can develop in an ignited inhomogeneous plasma. A complete analysis of the equations supporting the relevant theory is reported. Sponsored in part by the U.S. DoE.

NEUTRONIC REACTOR CONTROL ROD DRIVE APPARATUS

DOEpatents

Oakes, L.C.; Walker, C.S.

1959-12-15

ABS>A suspension mechanism between a vertically movable nuclear reactor control rod and a rod extension, which also provides information for the operator or an automatic control signal, is described. A spring connects the rod extension to a drive shift. The extension of the spring indicates whether (1) the rod is at rest on the reactor, (2) the rod and extension are suspended, or (3) the extension alone is suspended, the spring controlling a 3-position electrical switch.

NEUTRONIC REACTOR CONTROL ELEMENT

DOEpatents

Newson, H.W.

1960-09-13

A novel composite neutronic reactor control element is offered. The element comprises a multiplicity of sections arranged in end-to-end relationship, each of the sections having a markedly different neutron-reactive characteristic. For example, a three-section control element could contain absorber, moderator, and fuel sections. By moving such an element longitudinally through a reactor core, reactivity is decreased by the absorber, increased slightly by the moderator, or increased substantially by the fuel. Thus, control over a wide reactivity range is provided.

Split-core heat-pipe reactors for out-of-pile thermionic power systems.

NASA Technical Reports Server (NTRS)

Niederauer, G.; Lantz, E.; Breitweiser, R.

1971-01-01

Description of the concept of splitting a heat-pipe reactor for out-of-core thermionics into two identical halves and using the resulting center gap for reactivity control. Short Li-W reactor heat pipes penetrate the axial reflectors and form a heat exchanger with long heat pipes which wind through the shield to the thermionic diodes. With one reactor half anchored to the shield, the other is attached to a long arm with a pivot behind the shield and swings through a small arc for reactivity control. A safety shim prevents large reactivity inputs, and a fueled control arm drive shaft acts as a power stabilizer. Reactors fueled with U-235C and with U-233C have been studied.-

PRELIMINARY HAZARDS SUMMARY REPORT FOR THE VALLECITOS SUPERHEAT REACTOR

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

Murray, J.L.

1961-02-01

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

65. ARAII. Interior view of SL1 reactor building control piping ...

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

65. ARA-II. Interior view of SL-1 reactor building control piping for water purification system. On operating floor of building. March 21, 1958. Ineel photo no. 58-1360. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

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

Code of Federal Regulations, 2013 CFR

2013-01-01

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

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

77 FR 41670 - Definition of Terms

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-16

... cryptography'', 2. On page 642, add the term ``Explosives'', 3. On page 650, add the term ``Nuclear reactor... ``Commerce Control List''. * * * * * Nuclear reactor. (Cat 0 and 2) includes the items within or attached directly to the reactor vessel, the equipment which controls the level of power in the core, and the...

NEUTRONIC REACTOR MANIPULATING DEVICE

DOEpatents

Ohlinger, L.A.

1962-08-01

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Fission control system for nuclear reactor

DOEpatents

Conley, G.H.; Estes, G.P.

Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

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

REACTOR CONTROL ROD OPERATING SYSTEM

DOEpatents

Miller, G.

1961-12-12

A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 11 2011-07-01 2011-07-01 false Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

40 CFR 63.1406 - Reactor batch process vent provisions.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 11 2010-07-01 2010-07-01 true Reactor batch process vent provisions... § 63.1406 Reactor batch process vent provisions. (a) Emission standards. Owners or operators of reactor... reactor batch process vent located at a new affected source shall control organic HAP emissions by...

28. A typical main control panel in a 105 reactor ...

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

28. A typical main control panel in a 105 reactor building, in this case 105-F in February 1945. A single operator sat at the controls to regulate the pile's rate of reaction and monitor it for safety. The galvanometer screens (the two horizontal bars just below the nine round gauges that showed the positions of the control rods) showed the pile's current power setting. With that information, the operator could set the control rod positions to increase, decrease, or maintain the power. D-8310 - B Reactor, Richland, Benton County, WA

Safety control circuit for a neutronic reactor

DOEpatents

Ellsworth, Howard C.

2004-04-27

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

OVERALL CONTROL SYSTEM FOR HIGH FLUX PILE

DOEpatents

Newson, H.W.; Durham, N.C.; Wigner, E.P.; Princeton, N.J.; Epler, E.P.

1961-05-23

A control system is given for a high fiux reactor incorporating an anti- scram control feature whereby a neutron absorbing control rod acts as a fine adjustment while a neutron absorbing shim rod, actuated upon a command received from reactor period and level signals, has substantially greater effect on the neutron level and is moved prior to scram conditions to alter the reactor activity before a scram condition is created. Thus the probability that a scram will have to be initiated is substantially decreased.

Liquid phase methanol reactor staging process for the production of methanol

DOEpatents

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

1988-01-01

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

Control rod drive

DOEpatents

Hawke, Basil C.

1986-01-01

A control rod drive uses gravitational forces to insert one or more control rods upwardly into a reactor core from beneath the reactor core under emergency conditions. The preferred control rod drive includes a vertically movable weight and a mechanism operatively associating the weight with the control rod so that downward movement of the weight is translated into upward movement of the control rod. The preferred control rod drive further includes an electric motor for driving the control rods under normal conditions, an electrically actuated clutch which automatically disengages the motor during a power failure and a decelerator for bringing the control rod to a controlled stop when it is inserted under emergency conditions into a reactor core.

On a distinctive feature of problems of calculating time-average characteristics of nuclear reactor optimal control sets

NASA Astrophysics Data System (ADS)

Trifonenkov, A. V.; Trifonenkov, V. P.

2017-01-01

This article deals with a feature of problems of calculating time-average characteristics of nuclear reactor optimal control sets. The operation of a nuclear reactor during threatened period is considered. The optimal control search problem is analysed. The xenon poisoning causes limitations on the variety of statements of the problem of calculating time-average characteristics of a set of optimal reactor power off controls. The level of xenon poisoning is limited. There is a problem of choosing an appropriate segment of the time axis to ensure that optimal control problem is consistent. Two procedures of estimation of the duration of this segment are considered. Two estimations as functions of the xenon limitation were plot. Boundaries of the interval of averaging are defined more precisely.

Peculiarities of heat transfer at the liquid metal flow in a vertical channel in a coplanar magnetic field

NASA Astrophysics Data System (ADS)

Razuvanov, N. G.; Poddubnyi, I. I.; Kostychev, P. V.

2017-11-01

The research of hydrodynamics and heat transfer at the liquid metal (LM) downward flow and upflow in a vertical duct of a rectangular cross section with a ratio of sides ˜1/3 in a coplanar magnetic field (MF) under conditions of bilateral symmetrical heating is performed. The problem simulates the LM flow in the heat exchange channels for cooling the liquid metal module of the blanket of the thermonuclear reactor (TNR) of the TOKAMAK type. The experiments were carried out on the basis of the mercury magnetohydrodynamic test-bed (MHD) Moscow Power Engineering Institute (MPEI) - Joint Institute for High Temperatures of the Russian Academy of Sciences (JIHT RAS). The probe measurement technique was used in the flow. Profiles of averaged velocity and averaged temperature, as well as profiles of temperature pulsations in the axial planes of the channel cross-section, are obtained; the distribution of the dimensionless wall temperature along the perimeter unfolding of the channel in the section and along the length of the channel. A significant effect of thermogravitational convection (TGC), which leads to unexpected effects, is found. At the downflow in a magnetic field, in some modes, low-frequency pulsations of anomalously high intensity occur.

Dust dynamics and diagnostic applications in quasi-neutral plasmas and magnetic fusion

NASA Astrophysics Data System (ADS)

Wang, Zhehui; Ticos, Catalin M.; Si, Jiahe; Delzanno, Gian Luca; Lapenta, Gianni; Wurden, Glen

2007-11-01

Little is known about dust dynamics in highly ionized quasi-neutral plasmas with ca. 1.0 e+20 per cubic meter density and ion temperature at a few eV and above, including in magnetic fusion. For example, dust motion in fusion, better known as UFO's, has been observed since 1980's but not explained. Solid understanding of dust dynamics is also important to International Thermonuclear Experimental Reactor (ITER) because of concerns about safety and dust contamination of fusion core. Compared with well studied strongly-coupled dusty plasma regime, new physics may arise in the higher density quasi-neutral plasma regime because of at least four orders of magnitude higher density and two orders of magnitude hotter ion temperature. Our recent laboratory experiments showed that plasma-flow drag force dominates over other forces in a quasi-neutral flowing plasma. In contrast, delicate balance among different forces in dusty plasma has led to many unique phenomena, in particular, the formation of dust crystal. Based on our experiments, we argue that 1) dust crystal will not form in the highly ionized plasmas with flows; 2) the UFO's are moving dust dragged by plasma flows; 3) dust can be used to measure plasma flow. Two diagnostic applications using dust for laboratory quasi-neutral plasmas and magnetic fusion will also be presented.

Limitation of tritium outgassing from tritiated solid waste drums

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

Liger, K.; Trabuc, P.; Lefebvre, X.

2015-03-15

In the framework of the development of fusion thermonuclear reactors, tritiated solid waste is foreseen and will have to be managed. The management of tritiated waste implies limitations in terms of activity and tritium degassing. The degassing tritium can be under the form of tritiated hydrogen, tritiated water and, in some specific cases, negligible amount of tritiated volatile organic compound. Hence, considering the major forms of degassing tritium, CEA has developed a mixed-compound dedicated to tritium trapping in drums. Based on several experiments, the foreseen mixed compound is composed of MnO{sub 2}, Ag{sub 2}O, Pt and molecular sieve, the threemore » first species having the ability to convert tritiated hydrogen into tritiated water and the last one acting as a trap for tritiated water. To assess the performance of the trapping mixture, experimental tests were performed at room temperature on tritiated dust composed of beryllium and carbon. It was shown that the metallic oxides mixture used for tritiated hydrogen conversion is efficient and that tritiated water adsorption was limited due to an inefficient regeneration of the molecular sieve prior to its use. Apart from this point, the tritium release from waste was reduced by a factor of 5.5, which can be improved up to 87 if the adsorption step is efficient.« less

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

NASA Astrophysics Data System (ADS)

Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

2015-09-01

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

Numerical computation of gravitational field for general axisymmetric objects

NASA Astrophysics Data System (ADS)

Fukushima, Toshio

2016-10-01

We developed a numerical method to compute the gravitational field of a general axisymmetric object. The method (I) numerically evaluates a double integral of the ring potential by the split quadrature method using the double exponential rules, and (II) derives the acceleration vector by numerically differentiating the numerically integrated potential by Ridder's algorithm. Numerical comparison with the analytical solutions for a finite uniform spheroid and an infinitely extended object of the Miyamoto-Nagai density distribution confirmed the 13- and 11-digit accuracy of the potential and the acceleration vector computed by the method, respectively. By using the method, we present the gravitational potential contour map and/or the rotation curve of various axisymmetric objects: (I) finite uniform objects covering rhombic spindles and circular toroids, (II) infinitely extended spheroids including Sérsic and Navarro-Frenk-White spheroids, and (III) other axisymmetric objects such as an X/peanut-shaped object like NGC 128, a power-law disc with a central hole like the protoplanetary disc of TW Hya, and a tear-drop-shaped toroid like an axisymmetric equilibrium solution of plasma charge distribution in an International Thermonuclear Experimental Reactor-like tokamak. The method is directly applicable to the electrostatic field and will be easily extended for the magnetostatic field. The FORTRAN 90 programs of the new method and some test results are electronically available.

Cryogenic thermonuclear fuel implosions on the National Ignition Facility

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

Glenzer, S. H.; Callahan, D. A.; MacKinnon, A. J.

2012-05-15

The first inertial confinement fusion implosion experiments with equimolar deuterium-tritium thermonuclear fuel have been performed on the National Ignition Facility. These experiments use 0.17 mg of fuel with the potential for ignition and significant fusion yield conditions. The thermonuclear fuel has been fielded as a cryogenic layer on the inside of a spherical plastic capsule that is mounted in the center of a cylindrical gold hohlraum. Heating the hohlraum with 192 laser beams for a total laser energy of 1.6 MJ produces a soft x-ray field with 300 eV temperature. The ablation pressure produced by the radiation field compresses themore » initially 2.2-mm diameter capsule by a factor of 30 to a spherical dense fuel shell that surrounds a central hot-spot plasma of 50 {mu}m diameter. While an extensive set of x-ray and neutron diagnostics has been applied to characterize hot spot formation from the x-ray emission and 14.1 MeV deuterium-tritium primary fusion neutrons, thermonuclear fuel assembly is studied by measuring the down-scattered neutrons with energies in the range of 10 to 12 MeV. X-ray and neutron imaging of the compressed core and fuel indicate a fuel thickness of (14 {+-} 3) {mu}m, which combined with magnetic recoil spectrometer measurements of the fuel areal density of (1 {+-} 0.09) g cm{sup -2} result in fuel densities approaching 600 g cm{sup -3}. The fuel surrounds a hot-spot plasma with average ion temperatures of (3.5 {+-} 0.1) keV that is measured with neutron time of flight spectra. The hot-spot plasma produces a total fusion neutron yield of 10{sup 15} that is measured with the magnetic recoil spectrometer and nuclear activation diagnostics that indicate a 14.1 MeV yield of (7.5{+-}0.1) Multiplication-Sign 10{sup 14} which is 70% to 75% of the total fusion yield due to the high areal density. Gamma ray measurements provide the duration of nuclear activity of (170 {+-} 30) ps. These indirect-drive implosions result in the highest areal densities and neutron yields achieved on laser facilities to date. This achievement is the result of the first hohlraum and capsule tuning experiments where the stagnation pressures have been systematically increased by more than a factor of 10 by fielding low-entropy implosions through the control of radiation symmetry, small hot electron production, and proper shock timing. The stagnation pressure is above 100 Gbars resulting in high Lawson-type confinement parameters of P{tau} Asymptotically-Equal-To 10 atm s. Comparisons with radiation-hydrodynamic simulations indicate that the pressure is within a factor of three required for reaching ignition and high yield. This will be the focus of future higher-velocity implosions that will employ additional optimizations of hohlraum, capsule and laser pulse shape conditions.« less

The role and control of sludge age in biological nutrient removal activated sludge systems.

PubMed

Ekama, G A

2010-01-01

The sludge age is the most fundamental and important parameter in the design, operation and control of biological nutrient removal (BNR) activated sludge (AS) systems. Generally, the better the effluent and waste sludge quality required from the system, the longer the sludge age, the larger the biological reactor and the more wastewater characteristics need to be known. Controlling the reactor concentration does not control sludge age, only the mass of sludge in the system. When nitrification is a requirement, sludge age control becomes a requirement and the secondary settling tanks can no longer serve the dual purpose of clarifier and waste activated sludge thickeners. The easiest and most practical way to control sludge age is with hydraulic control by wasting a defined proportion of the reactor volume daily. In AS plants with reactor concentration control, nitrification fails first. With hydraulic control of sludge age, nitrification will not fail, rather the plant fails by shedding solids over the secondary settling tank effluent weirs.

A mixed helium-oxygen shell in some core-collapse supernova progenitors

NASA Astrophysics Data System (ADS)

Gofman, Roni Anna; Gilkis, Avishai; Soker, Noam

2018-04-01

We evolve models of rotating massive stars up to the stage of iron core collapse using the MESA code and find a shell with a mixed composition of primarily helium and oxygen in some cases. In the parameter space of initial masses of 13-40M⊙ and initial rotation velocities of 0-450 kms-1 that we investigate, we find a mixed helium-oxygen (He-O) shell with a significant total He-O mass and with a helium to oxygen mass ratio in the range of 0.5-2 only for a small fraction of the models. While the shell formation due to mixing is instigated by rotation, the pre-collapse rotation rate is not very high. The fraction of models with a shell of He-O composition required for an energetic collapse-induced thermonuclear explosion is small, as is the fraction of models with high specific angular momentum, which can aid the thermonuclear explosion by retarding the collapse. Our results suggest that the collapse-induced thermonuclear explosion mechanism that was revisited recently can account for at most a small fraction of core-collapse supernovae. The presence of such a mixed He-O shell still might have some implications for core-collapse supernovae, such as some nucleosynthesis processes when jets are present, or might result in peculiar sub-luminous core-collapse supernovae.

A mixed helium-oxygen shell in some core-collapse supernova progenitors

NASA Astrophysics Data System (ADS)

Gofman, Roni Anna; Gilkis, Avishai; Soker, Noam

2018-07-01

We evolve models of rotating massive stars up to the stage of iron core collapse using the MESA code and find a shell with a mixed composition of primarily helium and oxygen in some cases. In the parameter space of initial masses of 13-40 M⊙ and initial rotation velocities of 0-450 km s-1 that we investigate, we find a mixed helium-oxygen (He-O) shell with a significant total He-O mass and with a helium to oxygen mass ratio in the range of 0.5-2 only for a small fraction of the models. While the shell formation due to mixing is instigated by rotation, the pre-collapse rotation rate is not very high. The fraction of models with a shell of He-O composition required for an energetic collapse-induced thermonuclear explosion is small, as is the fraction of models with high specific angular momentum, which can aid the thermonuclear explosion by retarding the collapse. Our results suggest that the collapse-induced thermonuclear explosion mechanism that was revisited recently can account for at most a small fraction of core-collapse supernovae. The presence of such a mixed He-O shell still might have some implications for core-collapse supernovae, such as some nucleosynthesis processes when jets are present, or might result in peculiar sub-luminous core-collapse supernovae.

Cloud heights and stratospheric injections resulting from a thermonuclear war

NASA Astrophysics Data System (ADS)

Manins, P. C.

Two consequences of a major thermonuclear war are the injection of fireball material into the atmosphere and the production of vast quantities of dense smoke from fires which are ignited by the blasts. A major concern for assessment of impact on the environment is the height reached by this material. Fireball rise data are presented and a model for the plume rise from large fires in standard ambient conditions is validated with available data. It is concluded that injection of bomb debris into the stratosphere at mid and high latitudes should take place for all explosions with yield greater than approx. 30 kt of TNT equivalent. At low latitudes yields greater than 1 Mt are evidently required. Thus most fireball material would reach into the stratosphere under recently postulated scenarios. Fires would require a power output of 1.5 × 10 7 MW at middle and higher latitudes and 8 × 10 7 MW at low latitudes for significant injection of smoke into the stratosphere in standard conditions. Study of possible fires ignited in a thermonuclear war in rural and urban areas suggests that smoke from rural fires would reach the tropopause but that significant injections into the stratosphere are unlikely. Conflagration of large, medium- and high-density city-centres would, it is predicted, result in much smoke reaching to the tropopause and into the lower stratosphere at higher but not at low latitudes.

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG ...

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

MTR CONTROL ROOM WITH CONTROL CONSOLE AND STATUS READOUTS ALONG WALL. WORKERS MAKE ELECTRICAL AND OTHER CONNECTIONS. INL NEGATIVE NO. 4289. Unknown Photographer, 2/26/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

REACTOR

DOEpatents

Roman, W.G.

1961-06-27

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Reactor design and integration into a nuclear electric spacecraft

NASA Technical Reports Server (NTRS)

Phillips, W. M.; Koenig, D. R.

1978-01-01

One of the well-defined applications for nuclear power in space is nuclear electric propulsion (NEP). Mission studies have identified the optimum power level (400 kWe). A single Shuttle launch requirement and science-package integration have added additional constraints to the design. A reactor design which will meet these constraints has been studied. The reactor employs 90 fuel elements, each heat pipe cooled. Reactor control is obtained with BeO/B4C drums in a BeO reflector. The balance of the spacecraft is shielded from the reactor with LiH. Power conditioning and reactor control drum drives are located behind the LiH with the power conditioning. Launch safety, mechanical design and integration with the power conversion subsystem are discussed.

Analysis of a boron-carbide-drum-controlled critical reactor experiment

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

Fast-spectrum space-power-reactor concepts using boron control devices

NASA Technical Reports Server (NTRS)

Mayo, W.

1973-01-01

Several fast-spectrum space power reactor concepts that use boron carbide control devices were examined to determine the neutronic feasibility of the designs. The designs considered were (1) a 199-fuel-pin, 12-poison-reflector-control-drum reactor; (2) a 232-fuel-pin reactor with 12 reflector drums and three in-core control rods; (3) a 337-fuel-pin design with 12 incore control rods; and a 181-fuel-pin design with six drums closely coupled to the core to increase reactivity per drum. Adequate reactivity control and excess reactivity could be obtained for each concept, and the goals of 50,000 hours at 2.17 thermal megawatts with a lithium-7 coolant outlet temperature of 1222 K could be met without exceeding the 1-percent-clad-creep criterion. Heating rates in the boron carbide were calculated, but a heat transfer analysis was not done.

SPIRAL INSTABILITY CAN DRIVE THERMONUCLEAR EXPLOSIONS IN BINARY WHITE DWARF MERGERS

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

Kashyap, Rahul; Fisher, Robert; García-Berro, Enrique

2015-02-10

Thermonuclear, or Type Ia supernovae (SNe Ia), originate from the explosion of carbon–oxygen white dwarfs, and serve as standardizable cosmological candles. However, despite their importance, the nature of the progenitor systems that give rise to SNe Ia has not been hitherto elucidated. Observational evidence favors the double-degenerate channel in which merging white dwarf binaries lead to SNe Ia. Furthermore, significant discrepancies exist between observations and theory, and to date, there has been no self-consistent merger model that yields a SNe Ia. Here we show that a spiral mode instability in the accretion disk formed during a binary white dwarf mergermore » leads to a detonation on a dynamical timescale. This mechanism sheds light on how white dwarf mergers may frequently yield SNe Ia.« less

Fermi Establishes Classical Novae as a Distinct Class of Gamma-ray Sources

NASA Technical Reports Server (NTRS)

Ackermann, M.; Ajello, M.; Albert, A.; Baldini, L.; Ballet, J.; Bastieri, D.; Bellazzini, R.; Bissaldi, E.; Blandford, R. D.; Bloom, E. D.;

Fermi establishes classical novae as a distinct class of gamma-ray sources

DOE PAGES

Cheung, C. C.

2014-07-31

A classical nova results from runaway thermonuclear explosions on the surface of a white dwarf that accretes matter from a low-mass main-sequence stellar companion. In 2012 and 2013, three novae were detected in γ rays and stood in contrast to the first γ-ray detected nova V407 Cygni 2010, which belongs to a rare class of symbiotic binary systems. Despite likely differences in the compositions and masses of their white dwarf progenitors, the three classical novae are similarly characterized as soft spectrum transient γ-ray sources detected over 2-3 week durations. The γ-ray detections point to unexpected high-energy particle acceleration processes linkedmore » to the mass ejection from thermonuclear explosions in an unanticipated class of Galactic γ-ray sources.« less

Anticipatory control of xenon in a pressurized water reactor

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

Impink, A.J. Jr.

1987-02-10

A method is described for automatically dampening xenon-135 spatial transients in the core of a pressurized water reactor having control rods which regulate reactor power level, comprising the steps of: measuring the neutron flu in the reactor core at a plurality of axially spaced locations on a real-time, on-line basis; repetitively generating from the neutron flux measurements, on a point-by-point basis, signals representative of the current axial distribution of xenon-135, and signals representative of the current rate of change of the axial distribution of xenon-135; generating from the xenon-135 distribution signals and the rate of change of xenon distribution signals,more » control signals for reducing the xenon transients; and positioning the control rods as a function of the control signals to dampen the xenon-135 spatial transients.« less

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically includes the items within or attached directly to the reactor vessel, the equipment which controls the...

10 CFR Appendix A to Part 110 - Illustrative List of Nuclear Reactor Equipment Under NRC Export Licensing Authority

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Illustrative List of Nuclear Reactor Equipment Under NRC... List of Nuclear Reactor Equipment Under NRC Export Licensing Authority Note—A nuclear reactor basically includes the items within or attached directly to the reactor vessel, the equipment which controls the...

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

A novel plant protection system designed for use in the TREAT Upgrade (TU) reactor is described. The TU reactor is designed for controlled transient operation in the testing of reactor fuel behavior under simulated reactor accident conditions. Safe operation of the reactor is of paramount importance and the Plant Protection System (PPS) had to be designed to exacting requirements. Researchers believe that the strategy developed for the TU has potential application to the multimegawatt space reactors and represents the state of the art in terrestrial transient reactor protection systems.

Pressurized reactor system and a method of operating the same

DOEpatents

Isaksson, J.M.

1996-06-18

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Super-atmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gasification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor. 2 figs.

Pressurized reactor system and a method of operating the same

DOEpatents

Isaksson, Juhani M.

1996-01-01

A method and apparatus are provided for operating a pressurized reactor system in order to precisely control the temperature within a pressure vessel in order to minimize condensation of corrosive materials from gases on the surfaces of the pressure vessel or contained circulating fluidized bed reactor, and to prevent the temperature of the components from reaching a detrimentally high level, while at the same time allowing quick heating of the pressure vessel interior volume during start-up. Superatmospheric pressure gas is introduced from the first conduit into the fluidized bed reactor and heat derived reactions such as combustion and gassification are maintained in the reactor. Gas is exhausted from the reactor and pressure vessel through a second conduit. Gas is circulated from one part of the inside volume to another to control the temperature of the inside volume, such as by passing the gas through an exterior conduit which has a heat exchanger, control valve, blower and compressor associated therewith, or by causing natural convection flow of circulating gas within one or more generally vertically extending gas passages entirely within the pressure vessel (and containing heat exchangers, flow rate control valves, or the like therein). Preferably, inert gas is provided as a circulating gas, and the inert gas may also be used in emergency shut-down situations. In emergency shut-down reaction gas being supplied to the reactor is cut off, while inert gas from the interior gas volume of the pressure vessel is introduced into the reactor.

Influence of denitrification reactor retention time distribution (RTD) on dissolved oxygen control and nitrogen removal efficiency.

PubMed

Raboni, Massimo; Gavasci, Renato; Viotti, Paolo

2015-01-01

Low concentrations of dissolved oxygen (DO) are usually found in biological anoxic pre-denitrification reactors, causing a reduction in nitrogen removal efficiency. Therefore, the reduction of DO in such reactors is fundamental for achieving good nutrient removal. The article shows the results of an experimental study carried out to evaluate the effect of the anoxic reactor hydrodynamic model on both residual DO concentration and nitrogen removal efficiency. In particular, two hydrodynamic models were considered: the single completely mixed reactor and a series of four reactors that resemble plug-flow behaviour. The latter prove to be more effective in oxygen consumption, allowing a lower residual DO concentration than the former. The series of reactors also achieves better specific denitrification rates and higher denitrification efficiency. Moreover, the denitrification food to microrganism (F:M) ratio (F:MDEN) demonstrates a relevant synergic action in both controlling residual DO and improving the denitrification performance.

Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

NASA Astrophysics Data System (ADS)

Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

2010-07-01

Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Oxidative coupling of methane using inorganic membrane reactor

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

Ma, Y.H.; Moser, W.R.; Dixon, A.G.

1995-12-31

The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gasmore » phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.« less

Stimulation of the hydrolytic stage for biogas production from cattle manure in an electrochemical bioreactor.

PubMed

Samani, Saeed; Abdoli, Mohammad Ali; Karbassi, Abdolreza; Amin, Mohammad Mehdi

Electrical current in the hydrolytic phase of the biogas process might affect biogas yield. In this study, four 1,150 mL single membrane-less chamber electrochemical bioreactors, containing two parallel titanium plates were connected to the electrical source with voltages of 0, -0.5, -1 and -1.5 V, respectively. Reactor 1 with 0 V was considered as a control reactor. The trend of biogas production was precisely checked against pH, oxidation reduction potential and electrical power at a temperature of 37 ± 0.5°C amid cattle manure as substrate for 120 days. Biogas production increased by voltage applied to Reactors 2 and 3 when compared with the control reactor. In addition, the electricity in Reactors 2 and 3 caused more biogas production than Reactor 4. Acetogenic phase occurred more quickly in Reactor 3 than in the other reactors. The obtained results from Reactor 4 were indicative of acidogenic domination and its continuous behavior under electrical stimulation. The results of the present investigation clearly revealed that phasic electrical current could enhance the efficiency of biogas production.

Controlled CO preferential oxidation

DOEpatents

Meltser, Mark A.; Hoch, Martin M.

1997-01-01

Method for controlling the supply of air to a PROX reactor for the preferential oxidation in the presence of hydrogen wherein the concentration of the hydrogen entering and exiting the PROX reactor is monitored, the difference therebetween correlated to the amount of air needed to minimize such difference, and based thereon the air supply to the PROX reactor adjusted to provide such amount and minimize such difference.

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)

SNAP 10A FS-3 reactor performance

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

Hawley, J.P.; Johnson, R.A.

1966-08-15

SNAP 10FS-3 was the first flight-qualified SNAP reactor system to be operated in a simulated space environment. Prestart-up qualification testing, automatic start-up, endurance period performance, extended operation test and reactor shutdown are described as they affected, or were affected by, overall reactor performance. Performance of the reactor control system and the diagnostic instrumentation is critically evaluted.

New approach to control the methanogenic reactor of a two-phase anaerobic digestion system.

PubMed

von Sachs, Jürgen; Meyer, Ulrich; Rys, Paul; Feitkenhauer, Heiko

2003-03-01

A new control strategy for the methanogenic reactor of a two-phase anaerobic digestion system has been developed and successfully tested on the laboratory scale. The control strategy serves the purpose to detect inhibitory effects and to achieve good conversion. The concept is based on the idea that volatile fatty acids (VFA) can be measured in the influent of the methanogenic reactor by means of titration. Thus, information on the output (methane production) and input of the methanogenic reactor is available, and a (carbon) mass balance can be obtained. The control algorithm comprises a proportional/integral structure with the ratio of (a) the methane production rate measured online and (b) a maximum methane production rate expected (derived from the stoichiometry) as a control variable. The manipulated variable is the volumetric feed rate. Results are shown for an experiment with VFA (feed) concentration ramps and for experiments with sodium chloride as inhibitor.

Thermionic converter temperature controller

DOEpatents

Shaner, Benjamin J [McMurray, PA; Wolf, Joseph H [Pittsburgh, PA; Johnson, Robert G. R. [Trafford, PA

2001-04-24

A method and apparatus for controlling the temperature of a thermionic reactor over a wide range of operating power, including a thermionic reactor having a plurality of integral cesium reservoirs, a honeycomb material disposed about the reactor which has a plurality of separated cavities, a solid sheath disposed about the honeycomb material and having an opening therein communicating with the honeycomb material and cavities thereof, and a shell disposed about the sheath for creating a coolant annulus therewith so that the coolant in the annulus may fill the cavities and permit nucleate boiling during the operation of the reactor.

Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

NASA Technical Reports Server (NTRS)

Fox, T. A.

1973-01-01

An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

A study on using fireclay as a biomass carrier in an activated sludge system.

PubMed

Tilaki, Ramazan Ali Dianati

2011-01-01

By adding a biomass carrier to an activated sludge system, the biomass concentration will increase, and subsequently the organic removal efficiency will be enhanced. In this study, the possibility of using excess sludge from ceramic and tile manufacturing plants as a biomass carrier was investigated. The aim of this study was to determine the effect of using fireclay as a biomass carrier on biomass concentration, organic removal and nitrification efficiency in an activated sludge system. Experiments were conducted by using a bench scale activated sludge system operating in batch and continuous modes. Artificial simulated wastewater was made by using recirculated water in a ceramic manufacturing plant. In the continuous mode, hydraulic detention time in the aeration reactor was 8 and 22 h. In the batch mode, aeration time was 8 and 16 h. Fireclay doses were 500, 1,400 and 2,250 mg l(-1), and were added to the reactors in each experiment separately. The reactor with added fireclay was called a Hybrid Biological Reactor (HBR). A reactor without added fireclay was used as a control. Efficiency parameters such as COD, MLVSS and nitrate were measured in the control and HBR reactors according to standard methods. The average concentration of biomass in the HBR reactor was greater than in the control reactor. The total biomass concentration in the HBR reactor (2.25 g l(-1) fireclay) in the continuous mode was 3,000 mg l(-1) and in the batch mode was 2,400 mg l(-1). The attached biomass concentration in the HBR reactor (2.25 g l(-1) fireclay) in the continuous mode was 1,500 mg l(-1) and in the batch mode was 980 mg l(-1). Efficiency for COD removal in the HBR and control reactor was 95 and 55%, respectively. In the HBR reactor, nitrification was enhanced, and the concentration of nitrate was increased by 80%. By increasing the fireclay dose, total and attached biomass was increased. By adding fireclay as a biomass carrier, the efficiency of an activated sludge system to treat wastewater from ceramic manufacturing plants was increased.

Automated power control system for reactor TRIGA PUSPATI

NASA Astrophysics Data System (ADS)

Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

2017-01-01

Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

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

PubMed

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

2010-01-01

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

NUCLEAR REACTOR

DOEpatents

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

SERS internship Fall 1992--Spring 1993: Abstract and research papers

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

Not Available

1993-08-05

This report contains the abstracts and research papers by students on a variety of topics in engineering, genetics, solid state physics, thermonuclear energy, astrophysics, and other science related topics.

Model predictive control of a solar-thermal reactor

NASA Astrophysics Data System (ADS)

Saade Saade, Maria Elizabeth

Solar-thermal reactors represent a promising alternative to fossil fuels because they can harvest solar energy and transform it into storable and transportable fuels. The operation of solar-thermal reactors is restricted by the available sunlight and its inherently transient behavior, which affects the performance of the reactors and limits their efficiency. Before solar-thermal reactors can become commercially viable, they need to be able to maintain a continuous high-performance operation, even in the presence of passing clouds. A well-designed control system can preserve product quality and maintain stable product compositions, resulting in a more efficient and cost-effective operation, which can ultimately lead to scale-up and commercialization of solar thermochemical technologies. In this work, we propose a model predictive control (MPC) system for a solar-thermal reactor for the steam-gasification of biomass. The proposed controller aims at rejecting the disturbances in solar irradiation caused by the presence of clouds. A first-principles dynamic model of the process was developed. The model was used to study the dynamic responses of the process variables and to identify a linear time-invariant model used in the MPC algorithm. To provide an estimation of the disturbances for the control algorithm, a one-minute-ahead direct normal irradiance (DNI) predictor was developed. The proposed predictor utilizes information obtained through the analysis of sky images, in combination with current atmospheric measurements, to produce the DNI forecast. In the end, a robust controller was designed capable of rejecting disturbances within the operating region. Extensive simulation experiments showed that the controller outperforms a finely-tuned multi-loop feedback control strategy. The results obtained suggest that our controller is suitable for practical implementation.

CONTROL RODS FOR NUCLEAR REACTORS

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

Bell, F.R.

1963-01-16

A means for controlling the control rod in emergency, when it is desired to shutdown the reactor with the shortest possible delay, is described. When the emergency occurs the control rod is allowed to drop freely under gravity from the control rod support tube into the bore in the reactor core. A normal shutdown is reached almost at the lowest rod position. In the shut-down position and also below it, the control rod had its full effect of reducing the level of activity in the core. When the shut-down position was reached, a brake came into action to decelerate themore » rod and reduce shock and the likelihood of damage. (C.E.S.)« less

Magnetic switch for reactor control rod. [LMFBR

DOEpatents

Germer, J.H.

1982-09-30

A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

Magnetic switch for reactor control rod

DOEpatents

Germer, John H.

1986-01-01

A magnetic reed switch assembly for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electromagnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

CONTROL OF VOLATILE ORGANIC COMPOUNDS BY AN AC ENERGIZED FERROELECTRIC PELLET REACTOR AND A PULSED CORONA REACTOR

EPA Science Inventory

The paper gives results of a study to develop baseline engineering data to demonstrate the feasibility of application of plasma reactors to the destruction of various volatile organic compounds at ppm levels. Two laboratory-scale reactors, an alternating current energized ferroel...

A Semi-Batch Reactor Experiment for the Undergraduate Laboratory

ERIC Educational Resources Information Center

Derevjanik, Mario; Badri, Solmaz; Barat, Robert

2011-01-01

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

CONTROL RODS FOR NUCLEAR REACTOR CORES

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

Bell, F.R.

1961-11-15

A reactor control rod is designed which has increased effectiveness as compared with the width of the aperture in the pressure vessel through which it passes. The control rod carries six fins, three on each side, and two of the fins are fixed while the other, being adjustable, is capable of movement from between the fixed fins to an extended position. Thus, the control rod assembly can be arranged so that the parts within the core form a substantially complete shell around the reactor central axis, while the apertures on the pressure vessel wall are well spaced for strength. (D.L.C.)

Dynamic analysis of gas-core reactor system

NASA Technical Reports Server (NTRS)

Turner, K. H., Jr.

1973-01-01

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

SELF-REGULATING BOILING-WATER NUCLEAR REACTORS

DOEpatents

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

1960-08-16

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

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

Development and trial manufacturing of 1/2-scale partial mock-up of blanket box structure for fusion experimental reactor

NASA Astrophysics Data System (ADS)

Hashimoto, Toshiyuki; Takatsu, Hideyuki; Sato, Satoshi

1994-07-01

Conceptual design of breeding blanket has been discussed during the CDA (Conceptual Design Activities) of ITER (International Thermonuclear Experimental Reactor). Structural concept of breeding blanket is based on box structure integrated with first wall and shield, which consists of three coolant manifolds for first wall, breeding and shield regions. The first wall must have cooling channels to remove surface heat flux and nuclear heating. The box structure includes plates to form the manifolds and stiffening ribs to withstand enormous electromagnetic load, coolant pressure and blanket internal (purge gas) pressure. A 1/2-scale partial model of the blanket box structure for the outboard side module near midplane is manufactured to estimate the fabrication technology, i.e. diffusion bonding by HIP (Hot Isostatic Pressing) and EBW (Electron Beam Welding) procedure. Fabrication accuracy is a key issue to manufacture first wall panel because bending deformation during HIP may not be small for a large size structure. Data on bending deformation during HIP was obtained by preliminary manufacturing of HIP elements. For the shield structure, it is necessary to reduce the welding strain and residual stress of the weldment to establish the fabrication procedure. Optimal shape of the parts forming the manifolds, welding locations and welding sequence have been investigated. In addition, preliminary EBW tests have been performed in order to select the EBW conditions, and fundamental data on built-up shield have been obtained. Especially, welding deformation by joining the first wall panel to the shield has been measured, and total deformation to build-up shield by EBW has been found to be smaller than 2 mm. Consequently, the feasibility of fabrication technologies has been successfully demonstrated for a 1m-scaled box structure including the first wall with cooling channels by means of HIP, EBW and TIG (Tungsten Inert Gas arc)-welding.

Fusion materials: Technical evaluation of the technology of vandium alloys for use as blanket structural materials in fusion power systems

NASA Astrophysics Data System (ADS)

1993-08-01

The Committee's evaluation of vanadium alloys as a structural material for fusion reactors was constrained by limited data and time. The design of the International Thermonuclear Experimental Reactor is still in the concept stage, so meaningful design requirements were not available. The data on the effect of environment and irradiation on vanadium alloys were sparse, and interpolation of these data were made to select the V-5Cr-5Ti alloy. With an aggressive, fully funded program it is possible to qualify a vanadium alloy as the principal structural material for the ITER blanket in the available 5 to 8-year window. However, the data base for V-5Cr-5Ti is limited and will require an extensive development and test program. Because of the chemical reactivity of vanadium the alloy will be less tolerant of system failures, accidents, and off-normal events than most other candidate blanket structural materials and will require more careful handling during fabrication of hardware. Because of the cost of the material more stringent requirements on processes, and minimal historical working experience, it will cost an order of magnitude to qualify a vanadium alloy for ITER blanket structures than other candidate materials. The use of vanadium is difficult and uncertain; therefore, other options should be explored more thoroughly before a final selection of vanadium is confirmed. The Committee views the risk as being too high to rely solely on vanadium alloys. In viewing the state and nature of the design of the ITER blanket as presented to the Committee, it is obvious that there is a need to move toward integrating fabrication, welding, and materials engineers into the ITER design team. If the vanadium alloy option is to be pursued, a large program needs to be started immediately. The commitment of funding and other resources needs to be firm and consistent with a realistic program plan.

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

PubMed

Vasanthi, D; Pranavamoorthy, B; Pappa, N

2012-01-01

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

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

PubMed Central

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

2014-01-01

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

Advantages of Fast Ignition Scenarios with Two Hot Spots for Space Propulsion Systems

NASA Astrophysics Data System (ADS)

Shmatov, M. L.

The use of the fast ignition scenarios with the attempts to create two hot spots in one blob of the compressed thermonuclear fuel or, briefly, scenarios with two hot spots in space propulsion systems is proposed. The model, predicting that for such scenarios the probability pf of failure of ignition of thermonuclear microexplosion can be significantly less than that for the similar scenarios with the attempts to create one hot spot in one blob of the compressed fuel, is presented. For space propulsion systems consuming a relatively large amount of propellant, a decrease in pf due to the choice of the scenario with two hot spots can result in large, for example, two-fold, increase in the payload mass. Other advantages of the scenarios with two hot spots and some problems related to them are considered.

Gain curves and hydrodynamic modeling for shock ignition

NASA Astrophysics Data System (ADS)

Lafon, M.; Ribeyre, X.; Schurtz, G.

2010-05-01

Ignition of a precompressed thermonuclear fuel by means of a converging shock is now considered as a credible scheme to obtain high gains for inertial fusion energy. This work aims at modeling the successive stages of the fuel time history, from compression to final thermonuclear combustion, in order to provide the gain curves of shock ignition (SI). The leading physical mechanism at work in SI is pressure amplification, at first by spherical convergence, and by collision with the shock reflected at center during the stagnation process. These two effects are analyzed, and ignition conditions are provided as functions of the shock pressure and implosion velocity. Ignition conditions are obtained from a non-isobaric fuel assembly, for which we present a gain model. The corresponding gain curves exhibit a significantly lower ignition threshold and higher target gains than conventional central ignition.

NASA superconducting magnetic mirror facility. [for thermonuclear research

NASA Technical Reports Server (NTRS)

Reinmann, J. J.; Swanson, M. C.; Nichols, C. R.; Bloy, S. J.; Nagy, L. A.; Brady, F. J.

1973-01-01

The design details and initial test results of a superconducting magnetic mirror facility that has been constructed at NASA Lewis Research Center for use in thermonuclear research are summarized. The magnet system consists of four solenoidal coils which are individually rated at 5.0 T. Each coll is composed of an inner, middle, and outer winding. The inner winding is wound of stabilized Nb3Sn superconducting ribbon, and the middle and outer windings are wound of stabilized Nb-Ti superconducting wire. When arranged in the mirror geometry, the four coils will produce 8.7 T at the mirrors and a 1.8 mirror ratio. The magnet has a 41-cm diameter clear bore which is open to atmosphere. Distance between the mirrors is 111 cm. Presently there are only three magnets in the facility; the fourth magnet is being rebuilt.

Improving biogas production from continuous co-digestion of oily wastewater and waste-activated sludge by hydrodynamic cavitation pre-treatment.

PubMed

Habashi, Nima; Alighardashi, Abolghasem; Mennerich, Artur; Mehrdadi, Nasser; Torabian, Ali

2018-04-01

Hydrodynamic cavitation (HC) was evaluated as a pretreatment for synthetic oily wastewater (OWW) to be co-digested with waste-activated sludge (WAS). The main objective of the present research was the enhancement of biogas production by the application of HC pretreatment. HC was applied to the OWW, and the OWW and WAS were added to a 50 L continuous digestion reactor. As a control system, an identical digestion reactor was set up for co-digestion of the WAS and the OWW without pretreatment. The reactors were initially filled with inoculum and the hydraulic retention time (HRT) was set to 22 d. The HRT was gradually reduced to 19, 16, and finally 13 d, but the substrate quality was kept constant. The loading rate, accordingly, increased from 0.86 to 1.46 g TVS/(L d). The biogas volume was recorded online and its quality was analyzed regularly. The HC improved biogas production up to 43% at 22 d of HRT. Reducing the HRT decreased biogas production from the main reactor while that of the control reactor was more or less constant. HC also increased the biogas methane content; the methane concentration of the main reactor was about 3% higher than the methane concentration of the control reactor. The main reactor experienced no clogging or accumulation of fatty materials.

Nuclear reactor fuel containment safety structure

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

Rosewell, M.P.

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

A novel plant protection strategy for transient reactors

NASA Astrophysics Data System (ADS)

Bhattacharyya, Samit K.; Lipinski, Walter C.; Hanan, Nelson A.

The present plant protection system (PPS) has been defined for use in the TREAT-upgrade (TU) reactor for controlled transient operation of reactor-fuel behavior testing under simulated reactor-accident conditions. A PPS with energy-dependent trip set points lowered worst-case clad temperatures by as much as 180 K, relative to the use of conventional fixed-level trip set points. The multilayered multilevel protection strategy represents the state-of-the-art in terrestrial transient reactor protection systems, and should be applicable to multi-MW space reactors.

Nuclear Energy Policy

DTIC Science & Technology

2009-12-10

Small Modular Reactors Rising cost estimates for large conventional nuclear power plants—widely projected to be $6 billion or more—have contributed to growing interest in proposals for smaller, modular reactors. Ranging from about 40 to 350 megawatts of electrical capacity, such reactors would be only a fraction of the size of current commercial reactors. Several modular reactors would be installed together to make up a power block with a single control room, under most concepts. Modular reactor concepts would use a variety of technologies,

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

PubMed

Lozada, Mariana; Basile, Laura; Erijman, Leonardo

2007-01-01

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

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

Physical and chemical controls on the critical zone

USGS Publications Warehouse

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

2007-01-01

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

Nuclear reactor with low-level core coolant intake

DOEpatents

Challberg, Roy C.; Townsend, Harold E.

1993-01-01

A natural-circulation boiling-water reactor has skirts extending downward from control rod guide tubes to about 10 centimeters from the reactor vessel bottom. The skirts define annular channels about control rod drive housings that extend through the reactor vessel bottom. Recirculating water is forced in through the low-level entrances to these channels, sweeping bottom water into the channels in the process. The sweeping action prevents cooler water from accumulating at the bottom. This in turn minimizes thermal shock to bottom-dwelling components as would occur when accumulated cool water is swept away and suddenly replaced by warmer water.

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

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams.

On some control problems of dynamic of reactor

NASA Astrophysics Data System (ADS)

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

2017-12-01

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

Advanced gray rod control assembly

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

Drudy, Keith J; Carlson, William R; Conner, Michael E

An advanced gray rod control assembly (GRCA) for a nuclear reactor. The GRCA provides controlled insertion of gray rod assemblies into the reactor, thereby controlling the rate of power produced by the reactor and providing reactivity control at full power. Each gray rod assembly includes an elongated tubular member, a primary neutron-absorber disposed within the tubular member said neutron-absorber comprising an absorber material, preferably tungsten, having a 2200 m/s neutron absorption microscopic capture cross-section of from 10 to 30 barns. An internal support tube can be positioned between the primary absorber and the tubular member as a secondary absorber tomore » enhance neutron absorption, absorber depletion, assembly weight, and assembly heat transfer characteristics.« less

Analysis of space reactor system components: Investigation through simulation and non-nuclear testing

NASA Astrophysics Data System (ADS)

Bragg-Sitton, Shannon M.

The use of fission energy in space power and propulsion systems offers considerable advantages over chemical propulsion. Fission provides over six orders of magnitude higher energy density, which translates to higher vehicle specific impulse and lower specific mass. These characteristics enable ambitious space exploration missions. The natural space radiation environment provides an external source of protons and high energy, high Z particles that can result in the production of secondary neutrons through interactions in reactor structures. Applying the approximate proton source in geosynchronous orbit during a solar particle event, investigation using MCNPX 2.5.b for proton transport through the SAFE-400 heat pipe cooled reactor indicates an incoming secondary neutron current of (1.16 +/- 0.03) x 107 n/s at the core-reflector interface. This neutron current may affect reactor operation during low power maneuvers (e.g., start-up) and may provide a sufficient reactor start-up source. It is important that a reactor control system be designed to automatically adjust to changes in reactor power levels, maintaining nominal operation without user intervention. A robust, autonomous control system is developed and analyzed for application during reactor start-up, accounting for fluctuations in the radiation environment that result from changes in vehicle location or to temporal variations in the radiation field. Development of a nuclear reactor for space applications requires a significant amount of testing prior to deployment of a flight unit. High confidence in fission system performance can be obtained through relatively inexpensive non-nuclear tests performed in relevant environments, with the heat from nuclear fission simulated using electric resistance heaters. A series of non-nuclear experiments was performed to characterize various aspects of reactor operation. This work includes measurement of reactor core deformation due to material thermal expansion and implementation of a virtual reactivity feedback control loop; testing and thermal hydraulic characterization of the coolant flow paths for two space reactor concepts; and analysis of heat pipe operation during start-up and steady state operation.

CONTROL IN NUCLEAR REACTORS

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

Bell, F.R.

1963-01-16

An arrangement was described for scramming a reactor in an emergency. Control rods were position adjusted by an electric motor and transmission. A locking system kept the control rods in position but was arranged to be released in an emergency to allow the rods to drop into their shutdown position. (C.E.S.)

Autonomous Control of Space Reactor Systems

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

Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo

2007-11-30

Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

Electric-stepping-motor tests for a control-drum actuator of a nuclear reactor

NASA Technical Reports Server (NTRS)

Kieffer, A. W.

1972-01-01

Experimental tests were conducted on two stepping motors for application as reactor control-drum actuators. Various control-drum loads with frictional resistances ranging from approximately zero to 40 N-m and inertias ranging from zero to 0.424 kg-sq m were tested.

Self-actuating reactor shutdown system

DOEpatents

Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

1988-01-01

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

Adaptive Neural Network Algorithm for Power Control in Nuclear Power Plants

NASA Astrophysics Data System (ADS)

Masri Husam Fayiz, Al

2017-01-01

The aim of this paper is to design, test and evaluate a prototype of an adaptive neural network algorithm for the power controlling system of a nuclear power plant. The task of power control in nuclear reactors is one of the fundamental tasks in this field. Therefore, researches are constantly conducted to ameliorate the power reactor control process. Currently, in the Department of Automation in the National Research Nuclear University (NRNU) MEPhI, numerous studies are utilizing various methodologies of artificial intelligence (expert systems, neural networks, fuzzy systems and genetic algorithms) to enhance the performance, safety, efficiency and reliability of nuclear power plants. In particular, a study of an adaptive artificial intelligent power regulator in the control systems of nuclear power reactors is being undertaken to enhance performance and to minimize the output error of the Automatic Power Controller (APC) on the grounds of a multifunctional computer analyzer (simulator) of the Water-Water Energetic Reactor known as Vodo-Vodyanoi Energetichesky Reaktor (VVER) in Russian. In this paper, a block diagram of an adaptive reactor power controller was built on the basis of an intelligent control algorithm. When implementing intelligent neural network principles, it is possible to improve the quality and dynamic of any control system in accordance with the principles of adaptive control. It is common knowledge that an adaptive control system permits adjusting the controller’s parameters according to the transitions in the characteristics of the control object or external disturbances. In this project, it is demonstrated that the propitious options for an automatic power controller in nuclear power plants is a control system constructed on intelligent neural network algorithms.

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

DOEpatents

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Integrated intelligent systems in advanced reactor control rooms

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

Beckmeyer, R.R.

1989-01-01

An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs.,more » 5 figs.« less

Cost-Effective Systems for Atomic Layer Deposition

ERIC Educational Resources Information Center

Lubitz, Michael; Medina, Phillip A., IV; Antic, Aleks; Rosin, Joseph T.; Fahlman, Bradley D.

2014-01-01

Herein, we describe the design and testing of two different home-built atomic layer deposition (ALD) systems for the growth of thin films with sub-monolayer control over film thickness. The first reactor is a horizontally aligned hot-walled reactor with a vacuum purging system. The second reactor is a vertically aligned cold-walled reactor with a…

Operational performance of the three bean salad control algorithm on the ACRR (Annular Core Research Reactor)

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

Ball, R.M.; Madaras, J.J.; Trowbridge, F.R. Jr.

Experimental tests on the Annular Core Research Reactor have confirmed that the Three-Bean-Salad'' control algorithm based on the Pontryagin maximum principle can change the power of a nuclear reactor many decades with a very fast startup rate and minimal overshoot. The paper describes the results of simulations and operations up to 25 MW and 87 decades per minute. 3 refs., 4 figs., 1 tab.

Controlled CO preferential oxidation

DOEpatents

Meltser, M.A.; Hoch, M.M.

1997-06-10

Method is described for controlling the supply of air to a PROX (PReferential OXidation for CO cleanup) reactor for the preferential oxidation in the presence of hydrogen wherein the concentration of the hydrogen entering and exiting the PROX reactor is monitored, the difference there between correlated to the amount of air needed to minimize such difference, and based thereon the air supply to the PROX reactor adjusted to provide such amount and minimize such difference. 2 figs.

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION ...

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

MTR BASEMENT. GENERAL ELECTRIC CONTROL CONSOLE FOR AIRCRAFT NUCLEAR PROPULSION EXPERIMENT NO. 1. INL NEGATIVE NO. 6510. Unknown Photographer, 9/29/1959 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

Determining linear vibration frequencies of a ferromagnetic shell

NASA Astrophysics Data System (ADS)

Bagdoev, A. G.; Vardanyan, A. V.; Vardanyan, S. V.; Kukudzhanov, V. N.

2007-10-01

The problems of determining the roots of dispersion equations for free bending vibrations of thin magnetoelastic plates and shells are of both theoretical and practical interest, in particular, in studying vibrations of metallic structures used in controlled thermonuclear reactors. These problems were solved on the basis of the Kirchhoff hypothesis in [1-5]. In [6], an exact spatial approach to determining the vibration frequencies of thin plates was suggested, and it was shown that it completely agrees with the solution obtained according to the Kirchhoff hypothesis. In [7-9], this exact approach was used to solve the problem on vibrations of thin magnetoelastic plates, and it was shown by cumbersome calculations that the solutions obtained according to the exact theory and the Kirchhoff hypothesis differ substantially except in a single case. In [10], the equations of the dynamic theory of elasticity in the axisymmetric problem are given. In [11], the equations for the vibration frequencies of thin ferromagnetic plates with arbitrary conductivity were obtained in the exact statement. In [12], the Kirchhoff hypothesis was used to obtain dispersion relations for a magnetoelastic thin shell. In [5, 13-16], the relations for the Maxwell tensor and the ponderomotive force for magnetics were presented. In [17], the dispersion relations for thin ferromagnetic plates in the transverse field in the spatial statement were studied analytically and numerically. In the present paper, on the basis of the exact approach, we study free bending vibrations of a thin ferromagnetic cylindrical shell. We obtain the exact dispersion equation in the form of a sixth-order determinant, which can be solved numerically in the case of a magnetoelastic thin shell. The numerical results are presented in tables and compared with the results obtained by the Kirchhoff hypothesis. We show a large number of differences in the results, even for the least frequency.

In-vessel tritium retention and removal in ITER

NASA Astrophysics Data System (ADS)

Federici, G.; Anderl, R. A.; Andrew, P.; Brooks, J. N.; Causey, R. A.; Coad, J. P.; Cowgill, D.; Doerner, R. P.; Haasz, A. A.; Janeschitz, G.; Jacob, W.; Longhurst, G. R.; Nygren, R.; Peacock, A.; Pick, M. A.; Philipps, V.; Roth, J.; Skinner, C. H.; Wampler, W. R.

Tritium retention inside the vacuum vessel has emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER). In this paper we review recent tokamak and laboratory data on hydrogen, deuterium and tritium retention for materials and conditions which are of direct relevance to the design of ITER. These data, together with significant advances in understanding the underlying physics, provide the basis for modelling predictions of the tritium inventory in ITER. We present the derivation, and discuss the results, of current predictions both in terms of implantation and codeposition rates, and critically discuss their uncertainties and sensitivity to important design and operation parameters such as the plasma edge conditions, the surface temperature, the presence of mixed-materials, etc. These analyses are consistent with recent tokamak findings and show that codeposition of tritium occurs on the divertor surfaces primarily with carbon eroded from a limited area of the divertor near the strike zones. This issue remains an area of serious concern for ITER. The calculated codeposition rates for ITER are relatively high and the in-vessel tritium inventory limit could be reached, under worst assumptions, in approximately a week of continuous operation. We discuss the implications of these estimates on the design, operation and safety of ITER and present a strategy for resolving the issues. We conclude that as long as carbon is used in ITER - and more generically in any other next-step experimental fusion facility fuelled with tritium - the efficient control and removal of the codeposited tritium is essential. There is a critical need to develop and test in situ cleaning techniques and procedures that are beyond the current experience of present-day tokamaks. We review some of the principal methods that are being investigated and tested, in conjunction with the R&D work still required to extrapolate their applicability to ITER. Finally, unresolved issues are identified and recommendations are made on potential R&D avenues for their resolution.

High Temperature Uniaxial Compression and Stress-Relaxation Behavior of India-Specific RAFM Steel

NASA Astrophysics Data System (ADS)

Shah, Naimish S.; Sunil, Saurav; Sarkar, Apu

2018-07-01

India-specific reduced activity ferritic martensitic steel (INRAFM), a modified 9Cr-1Mo grade, has been developed by India as its own structural material for fabrication of the Indian Test Blanket Module (TBM) to be installed in the International Thermonuclear Energy Reactor (ITER). The extensive study on mechanical and physical properties of this material has been currently going on for appraisal of this material before being put to use in the ITER. High temperature compression, stress-relaxation, and strain-rate change behavior of the INRAFM steel have been investigated. The optical microscopic and scanning electron microscopic characterizations were carried out to observe the microstructural changes that occur during uniaxial compressive deformation test. Comparable true plastic stress values at 300 °C and 500 °C and a high drop in true plastic stress at 600 °C were observed during the compression test. Stress-relaxation behaviors were investigated at 500 °C, 550 °C, and 600 °C at a strain rate of 10-3 s-1. The creep properties of the steel at different temperatures were predicted from the stress-relaxation test. The Norton's stress exponent ( n) was found to decrease with the increasing temperature. Using Bird-Mukherjee-Dorn relationship, the temperature-compensated normalized strain rate vs stress was plotted. The stress exponent ( n) value of 10.05 was obtained from the normalized plot. The increasing nature of the strain rate sensitivity ( m) with the test temperature was found from strain-rate change test. The low plastic stability with m 0.06 was observed at 600 °C. The activation volume ( V *) values were obtained in the range of 100 to 300 b3. By comparing the experimental values with the literature, the rate-controlling mechanisms at the thermally activated region of high temperature were found to be the nonconservative movement of jogged screw dislocations and thermal breaking of attractive junctions.

High Temperature Uniaxial Compression and Stress-Relaxation Behavior of India-Specific RAFM Steel

NASA Astrophysics Data System (ADS)

Shah, Naimish S.; Sunil, Saurav; Sarkar, Apu

2018-05-01

India-specific reduced activity ferritic martensitic steel (INRAFM), a modified 9Cr-1Mo grade, has been developed by India as its own structural material for fabrication of the Indian Test Blanket Module (TBM) to be installed in the International Thermonuclear Energy Reactor (ITER). The extensive study on mechanical and physical properties of this material has been currently going on for appraisal of this material before being put to use in the ITER. High temperature compression, stress-relaxation, and strain-rate change behavior of the INRAFM steel have been investigated. The optical microscopic and scanning electron microscopic characterizations were carried out to observe the microstructural changes that occur during uniaxial compressive deformation test. Comparable true plastic stress values at 300 °C and 500 °C and a high drop in true plastic stress at 600 °C were observed during the compression test. Stress-relaxation behaviors were investigated at 500 °C, 550 °C, and 600 °C at a strain rate of 10-3 s-1. The creep properties of the steel at different temperatures were predicted from the stress-relaxation test. The Norton's stress exponent (n) was found to decrease with the increasing temperature. Using Bird-Mukherjee-Dorn relationship, the temperature-compensated normalized strain rate vs stress was plotted. The stress exponent (n) value of 10.05 was obtained from the normalized plot. The increasing nature of the strain rate sensitivity (m) with the test temperature was found from strain-rate change test. The low plastic stability with m 0.06 was observed at 600 °C. The activation volume (V *) values were obtained in the range of 100 to 300 b3. By comparing the experimental values with the literature, the rate-controlling mechanisms at the thermally activated region of high temperature were found to be the nonconservative movement of jogged screw dislocations and thermal breaking of attractive junctions.

Self isolating high frequency saturable reactor

DOEpatents

Moore, James A.

1998-06-23

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

Microprocessor tester for the treat upgrade reactor trip system

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

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety systemmore » is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations.« less

Pellet injector development at ORNL (Oak Ridge National Laboratory)

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Argo, B. E.; Baylor, L. R.; Combs, S. K.; Fehling, D. T.; Fisher, P. W.; Foster, C. A.; Foust, C. R.; Milora, S. L.; Qualls, A. L.

1990-09-01

Advanced plasma fueling systems for magnetic confinement experiments are under development at Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets to speeds in the kilometer-per-second range by either pneumatic (light-gas gun) or mechanical (centrifugal force) techniques. ORNL has recently provided a centrifugal pellet injector for the Tore Supra tokamak and a new, simplified, eight-shot pneumatic injector for the Advanced Toroidal Facility stellarator at ORNL. Hundreds of tritium and DT pellets were accelerated at the Tritium Systems Test Assembly facility at Los Alamos in 1988 to 1989. These experiments, done in a single-shot pipe-gun system, demonstrated the feasibility of forming and accelerating tritium pellets at low (sup 3)He levels. A new, tritium-compatible extruder mechanism is being designed for longer-pulse DT applications. Two-stage light-gas guns and electron beam rocket accelerators for speeds of the order of 2 to 10 km/s are also under development. Recently, a repeating, two-stage light-gas gun accelerated 10 surrogate pellets at a 1-Hz repetition rate to speeds in the range of 2 to 3 km/s; and the electron beam rocket accelerator completed initial feasibility and scaling experiments. ORNL has also developed conceptual designs of advanced plasma fueling systems for the Compact Ignition Tokamak and the International Thermonuclear Experimental Reactor.

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

Harvey, R.W.; Nevins, W.M.; Smith, G.R.

Electron cyclotron (EC) power has technological and physics advantages for heating and current drive in a tokamak reactor, and advances in source development make it credible for applications in ITER. Strong single pass absorption makes heating to ignition particularly simple. The optimized EC current drive (ECCD) efficiency ({l_angle}n{r_angle}IR/P) shows a linear temperature scaling at temperatures up to {approximately} 15 keV. For temperatures above 30 keV, the efficiency saturates at approximately 0.3{center_dot}10{sup 20} A/(m{sup 2}W) for a frequency of 220 GHz in an ITER target plasma with toroidal field of 6 T, due primarily to harmonic overlap [G.R. Smith et al.,more » Phys. Fluids 30 3633 (1987)] and to a lesser extent due to limitations arising from relativistic effects [N.J. Fisch, Phys. Rev. A 24 3245 (1981)]. The same efficiency can also be obtained at 170 GHz for the same plasma equilibrium except that the magnetic field is reduced to (170/220) {times} 6 T = 4.6 T. The ECCD efficiencies are obtained with the comprehensive 3D, bounce-averaged Fokker-Planck CQL3D codes [R.W. Harvey and M.G. McCoy, Proc. IAEA TCM/Advances in Simulation and Modeling in Thermonuclear Plasmas 1992, Montreal], and BANDIT3D [M.R. O`Brien, M. Cox, C.D. Warrick, and F. S. Zaitsev, ibid.].« less

Conceptual Design and Analysis of Cold Mass Support of the CS3U Feeder for the ITER

NASA Astrophysics Data System (ADS)

Zhu, Yinfeng; Song, Yuntao; Zhang, Yuanbin; Wang, Zhongwei

2013-06-01

In the International Thermonuclear Experimental Reactor (ITER) project, the feeders are one of the most important and critical systems. To convey the power supply and the coolant for the central solenoid (CS) magnet, 6 sets of CS feeders are employed, which consist mainly of an in-cryostat feeder (ICF), a cryostat feed-through (CFT), an S-bend box (SBB), and a coil terminal box (CTB). To compensate the displacements of the internal components of the CS feeders during operation, sliding cold mass supports consisting of a sled plate, a cylindrical support, a thermal shield, and an external ring are developed. To check the strength of the developed cold mass supports of the CS3U feeder, electromagnetic analysis of the two superconducting busbars is performed by using the CATIA V5 and ANSYS codes based on parametric technology. Furthermore, the thermal-structural coupling analysis is performed based on the obtained results, except for the stress concentration, and the max. stress intensity is lower than the allowable stress of the selected material. It is found that the conceptual design of the cold mass support can satisfy the required functions under the worst case of normal working conditions. All these performed activities will provide a firm technical basis for the engineering design and development of cold mass supports.

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

Zhai, Y.; D'Hauthuille, L.; Barth, C.

High-field superconducting magnets play a very important role in many large-scale physics experiments, particularly particle colliders and fusion confinement devices such as Large Hadron Collider (LHC) and International Thermonuclear Experimental Reactor (ITER). The two most common superconductors used in these applications are NbTi and Nb 3Sn. Nb 3Sn wires are favored because of their significantly higher J c (critical current density) for higher field applications. The main disadvantage of Nb 3Sn is that the superconducting performance of the wire is highly strain sensitive and it is very brittle. This strain sensitivity is strongly influenced by two factors: plasticity and crackedmore » filaments. Cracks are induced by large stress concentrators that can be traced to the presence of voids in the wire. We develop detailed 2-D and 3-D finite-element models containing wire filaments and different possible distributions of voids in a bronze-route Nb 3Sn wire. We apply compressive transverse loads for various cases of void distributions to simulate the stress and strain response of a Nb 3Sn wire under the Lorentz force. Furthermore, this paper improves our understanding of the effect voids have on the Nb 3Sn wire's mechanical properties, and in so, the connection between the distribution of voids and performance degradation such as the correlation between irreversible strain limit and the void-induced local stress concentrations.« less

Students' Knowledge of Nuclear Science and Its Connection with Civic Scientific Literacy in Two European Contexts: The Case of Newspaper Articles

NASA Astrophysics Data System (ADS)

Tsaparlis, Georgios; Hartzavalos, Sotiris; Nakiboğlu, Canan

2013-08-01

Nuclear science has uses and applications that are relevant and crucial for world peace and sustainable development, so knowledge of its basic concepts and topics should constitute an integral part of civic scientific literacy. We have used two newspaper articles that deal with uses of nuclear science that are directly relevant to life, society, economy, and international politics. One article discusses a new thermonuclear reactor, and the second one is about depleted uranium and its danger for health. 189 first-year undergraduate physics and primary education Greek students were given one of the two articles each, and asked to answer a number of accompanying questions dealing with knowledge that is part of the Greek high school curriculum. The study was repeated with 272 first-year undergraduate physics, physics education, science education, and primary education Turkish students. Acceptable or partially acceptable answers were provided on average by around 20 % of Greek and 11 % of Turkish students, while a large proportion (on the average, around 50 % of Greek and 27 % of Turkish students) abstained from answering the questions. These findings are disappointing, but should be seen in the light of the limited or no coverage of the relevant learning material in the Greek and the Turkish high-school programs. Student conceptual difficulties, misconceptions and implications for research and high school curricula are discussed.

Design Performance of Front Steering-Type Electron Cyclotron Launcher for ITER

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

Takahashi, K.; Imai, T.; Kobayashi, N.

2005-01-15

The performance of a front steering (FS)-type electron cyclotron launcher designed for the International Thermonuclear Experimental Reactor (ITER) is evaluated with a thermal, electromagnetic, and nuclear analysis of the components; a mechanical test of a spiral tube for the steering mirror; and a rotational test of bearings. The launcher consists of a front shield and a launcher plug where three movable optic mirrors to steer incident multimegawatt radio-frequency beam power, waveguide components, nuclear shields, and vacuum windows are installed. The windows are located behind a closure plate to isolate the transmission lines from the radioactivated circumstance (vacuum vessel). The waveguidemore » lines of the launcher are doglegged to reduce the direct neutron streaming toward the vacuum windows and other components. The maximum stresses on the critical components such as the steering mirror, its cooling tube, and the front shield are less than their allowable stresses. It was also identified that the stress on the launcher, which yielded from electromagnetic force caused by plasma disruption, was a little larger than the criteria, and a modification of the launcher plug structure was necessary. The nuclear analysis result shows that the neutron shield capability of the launcher satisfies the shield criteria of the ITER. It concludes that the design of the FS launcher is generally suitable for application to the ITER.« less

Homopolar machine for reversible energy storage and transfer systems

DOEpatents

Stillwagon, Roy E.

1978-01-01

A homopolar machine designed to operate as a generator and motor in reversibly storing and transferring energy between the machine and a magnetic load coil for a thermo-nuclear reactor. The machine rotor comprises hollow thin-walled cylinders or sleeves which form the basis of the system by utilizing substantially all of the rotor mass as a conductor thus making it possible to transfer substantially all the rotor kinetic energy electrically to the load coil in a highly economical and efficient manner. The rotor is divided into multiple separate cylinders or sleeves of modular design, connected in series and arranged to rotate in opposite directions but maintain the supply of current in a single direction to the machine terminals. A stator concentrically disposed around the sleeves consists of a hollow cylinder having a number of excitation coils each located radially outward from the ends of adjacent sleeves. Current collected at an end of each sleeve by sleeve slip rings and brushes is transferred through terminals to the magnetic load coil. Thereafter, electrical energy returned from the coil then flows through the machine which causes the sleeves to motor up to the desired speed in preparation for repetition of the cycle. To eliminate drag on the rotor between current pulses, the brush rigging is designed to lift brushes from all slip rings in the machine.

Homopolar machine for reversible energy storage and transfer systems

DOEpatents

Stillwagon, Roy E.

1981-01-01

A homopolar machine designed to operate as a generator and motor in reversibly storing and transferring energy between the machine and a magnetic load coil for a thermo-nuclear reactor. The machine rotor comprises hollow thin-walled cylinders or sleeves which form the basis of the system by utilizing substantially all of the rotor mass as a conductor thus making it possible to transfer substantially all the rotor kinetic energy electrically to the load coil in a highly economical and efficient manner. The rotor is divided into multiple separate cylinders or sleeves of modular design, connected in series and arranged to rotate in opposite directions but maintain the supply of current in a single direction to the machine terminals. A stator concentrically disposed around the sleeves consists of a hollow cylinder having a number of excitation coils each located radially outward from the ends of adjacent sleeves. Current collected at an end of each sleeve by sleeve slip rings and brushes is transferred through terminals to the magnetic load coil. Thereafter, electrical energy returned from the coil then flows through the machine which causes the sleeves to motor up to the desired speed in preparation for repetition of the cycle. To eliminate drag on the rotor between current pulses, the brush rigging is designed to lift brushes from all slip rings in the machine.

The effect of relativistic Compton scattering on thermonuclear burn of pure deuterium fuel

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

Ghasemizad, A.; Nazirzadeh, M.; Khanbabaei, B.

The relativistic effects of the Compton scattering on the thermonuclear burn-up of pure deuterium fuel in non-equilibrium plasma have been studied by four temperature (4T) theory. In the limit of low electron temperatures and photon energies, the nonrelativistic Compton scattering is valid and a convenient approximation, but in the high energy exchange rates between electrons and photons, is seen to break down. The deficiencies of the nonrelativistic approximation can be overcome by using the relativistic correction in the photons kinetic equation. In this research, we have utilized the four temperature (4T) theory to calculate the critical burn-up parameter for puremore » deuterium fuel, while the Compton scattering is considered as a relativistic phenomenon. It was shown that the measured critical burn-up parameter in ignition with relativistic Compton scattering is smaller than that of the parameter in the ignition with the nonrelativistic Compton scattering.« less

ON THE EFFECT OF EXPLOSIVE THERMONUCLEAR BURNING ON THE ACCRETED ENVELOPES OF WHITE DWARFS IN CATACLYSMIC VARIABLES

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

Sion, Edward M.; Sparks, Warren, E-mail: edward.sion@villanova.edu, E-mail: warrensparks@comcast.net

2014-11-20

The detection of heavy elements at suprasolar abundances in the atmospheres of some accreting white dwarfs in cataclysmic variables (CVs), coupled with the high temperatures needed to produce these elements, requires explosive thermonuclear burning. The central temperatures of any formerly more massive secondary stars in CVs undergoing hydrostatic CNO burning are far too low to produce these elements. Evidence is presented that at least some CVs contain donor secondaries that have been contaminated by white dwarf remnant burning during the common envelope phase and are transferring this material back to the white dwarf. This scenario does not exclude the channelmore » in which formerly more massive donor stars underwent CNO processing in systems with thermal timescale mass transfer. Implications for the progenitors of CVs are discussed and a new scenario for the white dwarf's accretion-nova-outburst is given.« less

The Dynamic Mutation Characteristics of Thermonuclear Reaction in Tokamak

PubMed Central

Li, Jing; Quan, Tingting; Zhang, Wei; Deng, Wei

2014-01-01

The stability and bifurcations of multiple limit cycles for the physical model of thermonuclear reaction in Tokamak are investigated in this paper. The one-dimensional Ginzburg-Landau type perturbed diffusion equations for the density of the plasma and the radial electric field near the plasma edge in Tokamak are established. First, the equations are transformed to the average equations with the method of multiple scales and the average equations turn to be a Z 2-symmetric perturbed polynomial Hamiltonian system of degree 5. Then, with the bifurcations theory and method of detection function, the qualitative behavior of the unperturbed system and the number of the limit cycles of the perturbed system for certain groups of parameter are analyzed. At last, the stability of the limit cycles is studied and the physical meaning of Tokamak equations under these parameter groups is given. PMID:24892099

Diffuse Galactic antimatter from faint thermonuclear supernovae in old stellar populations

NASA Astrophysics Data System (ADS)

Crocker, Roland M.; Ruiter, Ashley J.; Seitenzahl, Ivo R.; Panther, Fiona H.; Sim, Stuart; Baumgardt, Holger; Möller, Anais; Nataf, David M.; Ferrario, Lilia; Eldridge, J. J.; White, Martin; Tucker, Brad E.; Aharonian, Felix

2017-06-01

Our Galaxy hosts the annihilation of a few 1043 low-energy positrons every second. Radioactive isotopes capable of supplying such positrons are synthesized in stars, stellar remnants and supernovae. For decades, however, there has been no positive identification of a main stellar positron source, leading to suggestions that many positrons originate from exotic sources like the Galaxy's central supermassive black hole or dark matter annihilation. Here we show that a single type of transient source, deriving from stellar populations of age 3-6 Gyr and yielding ∼0.03 M ⊙ of the positron emitter 44Ti, can simultaneously explain the strength and morphology of the Galactic positron annihilation signal and the Solar System abundance of the 44Ti decay product 44Ca. This transient is likely the merger of two low-mass white dwarfs, observed in external galaxies as the sub-luminous, thermonuclear supernova known as SN 1991bg-like.

Burst Oscillation Probes of Neutron Stars and Nuclear Burning with LOFT

NASA Technical Reports Server (NTRS)

Strohmayer, Tod

2012-01-01

X-ray brightness oscillations during thermonuclear X-ray bursts--burst oscillations--have provided a new probe of neutron star spins as well as of the dependent nuclear burning processes. The frequency drift and amplitude evolution of the oscillations observed during bursts can in principle place constraints on the physics of thermonuclear flame spreading and the dynamics of the burning atmosphere. I use simulations appropriate to LOFT to explore the precision with which the time dependence of the oscillation frequency can be inferred. This can test, for example, different models for the frequency drift, such as up-lift versus geostrophic drift. I also explore the precision with which asymptotic frequencies can be constrained in order to estimate the capability for LOFT to detect the Doppler shifts induced by orbital motion of the neutron star from a sample of bursts at different orbital phases.