External events analysis for experimental fusion facilities

International Nuclear Information System (INIS)

Cadwallader, L.C.

1990-01-01

External events are those off-normal events that threaten facilities either from outside or inside the building. These events, such as floods, fires, and earthquakes, are among the leading risk contributors for fission power plants, and the nature of fusion facilities indicates that they may also lead fusion risk. This paper gives overviews of analysis methods, references good analysis guidance documents, and gives design tips for mitigating the effects of floods and fires, seismic events, and aircraft impacts. Implications for future fusion facility siting are also discussed. Sites similar to fission plant sites are recommended. 46 refs

The materials production and processing facility at the Spanish National Centre for fusion technologies (TechnoFusion)

International Nuclear Information System (INIS)

Munoz, A.; Monge, M.A.; Pareja, R.; Hernandez, M.T.; Jimenez-Rey, D.; Roman, R.; Gonzalez, M.; Garcia-Cortes, I.; Perlado, M.; Ibarra, A.

2011-01-01

In response to the urgent request from the EU Fusion Program, a new facility (TechnoFusion) for research and development of fusion materials has been planned with support from the Regional Government of Madrid and the Ministry of Science and Innovation of Spain. TechnoFusion, the National Centre for Fusion Technologies, aims screening different technologies relevant for ITER and DEMO environments while promoting the contribution of international companies and research groups into the Fusion Programme. For this purpose, the centre will be provided with a large number of unique facilities for the manufacture, testing (a triple-beam multi-ion irradiation, a plasma-wall interaction device, a remote handling for under ionizing radiation testing) and analysis of critical fusion materials. Particularly, the objectives, semi-industrial scale capabilities and present status of the TechnoFusion Materials Production and Processing (MPP) facility are presented. Previous studies revealed that the MPP facility will be a very promising infrastructure for the development of new materials and prototypes demanded by the fusion technology and therefore some of them will be here briefly summarized.

The materials production and processing facility at the Spanish National Centre for fusion technologies (TechnoFusion)

Energy Technology Data Exchange (ETDEWEB)

Munoz, A., E-mail: rpp@fis.uc3m.es [Departamento de Fisica, UC3M, Avda de la Universidad 30, 28911 Leganes, Madrid (Spain); Monge, M.A.; Pareja, R. [Departamento de Fisica, UC3M, Avda de la Universidad 30, 28911 Leganes, Madrid (Spain); Hernandez, M.T. [LNF-CIEMAT, Avda, Complutense, 22, 28040 Madrid (Spain); Jimenez-Rey, D. [CMAM, UAM, C/Faraday 3, 28049, Madrid (Spain); Roman, R.; Gonzalez, M.; Garcia-Cortes, I. [LNF-CIEMAT, Avda, Complutense, 22, 28040 Madrid (Spain); Perlado, M. [IFN, ETSII, UPM, C/Jose Gutierrez Abascal, 2, 28006 Madrid (Spain); Ibarra, A. [LNF-CIEMAT, Avda, Complutense, 22, 28040 Madrid (Spain)

2011-10-15

In response to the urgent request from the EU Fusion Program, a new facility (TechnoFusion) for research and development of fusion materials has been planned with support from the Regional Government of Madrid and the Ministry of Science and Innovation of Spain. TechnoFusion, the National Centre for Fusion Technologies, aims screening different technologies relevant for ITER and DEMO environments while promoting the contribution of international companies and research groups into the Fusion Programme. For this purpose, the centre will be provided with a large number of unique facilities for the manufacture, testing (a triple-beam multi-ion irradiation, a plasma-wall interaction device, a remote handling for under ionizing radiation testing) and analysis of critical fusion materials. Particularly, the objectives, semi-industrial scale capabilities and present status of the TechnoFusion Materials Production and Processing (MPP) facility are presented. Previous studies revealed that the MPP facility will be a very promising infrastructure for the development of new materials and prototypes demanded by the fusion technology and therefore some of them will be here briefly summarized.

Safety of magnetic fusion facilities: Requirements

International Nuclear Information System (INIS)

1996-05-01

This Standard identifies safety requirements for magnetic fusion facilities. Safety functions are used to define outcomes that must be achieved to ensure that exposures to radiation, hazardous materials, or other hazards are maintained within acceptable limits. Requirements applicable to magnetic fusion facilities have been derived from Federal law, policy, and other documents. In addition to specific safety requirements, broad direction is given in the form of safety principles that are to be implemented and within which safety can be achieved

Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

International Nuclear Information System (INIS)

Klueh, R.L.; Bloom, E.E.

1981-12-01

Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed

Development of fusion first-wall radiation damage facilities

International Nuclear Information System (INIS)

McElroy, R.J.; Atkins, T.

1986-11-01

The report describes work performed on the development of fusion-reactor first-wall simulation facilities on the Variable Energy Cyclotron, at Harwell, United Kingdom. Two irradiation facilities have been constructed: i) a device for helium and hydrogen filling up to 1000 ppm for post-irradiation mechanical properties studies, and ii) a helium implantation and damage facility for simultaneous injection of helium and radiation damage into a specimen under stress. These facilities are now fully commissioned and are available for investigations of first-wall radiation damage and for intercorrelation of fission- and fusion -reactor materials behaviour. (U.K.)

Survey of tritium wastes and effluents in near-term fusion-research facilities

International Nuclear Information System (INIS)

Bickford, W.E.; Dingee, D.A.; Willingham, C.E.

1981-08-01

The use of tritium control technology in near-term research facilities has been studied for both the magnetic and inertial confinement fusion programs. This study focused on routine generation of tritium wastes and effluents, with little referene to accidents or facility decommissioning. This report serves as an independent review of the effectiveness of planned control technology and radiological hazards associated with operation. The facilities examined for the magnetic fusion program included Fusion Materials Irradiation Testing Facility (FMIT), Tritium Systems Test Assembly (TSTA), and Tokamak Fusion Test Reactor (TFTR) in the magnetic fusion program, while NOVA and Antares facilities were examined for the inertial confinement program

Tritium handling facility at KMS Fusion Inc

International Nuclear Information System (INIS)

Bowman, C.C.; Vis, V.A.

1990-01-01

The tritium facility at KMS Fusion, Inc. supports the inertial confinement fusion research program. The main function of the facility is to fill glass and polymer Microshell (TM) capsules (small fuel containers) to a maximum pressure of 100 atm with tritium (T 2 ) or deuterium--tritium (DT). The recent upgrade of the facility allows us to fill Microshell capsules to a maximum pressure of 200 atm. A second fill port allows us to run long term fills of Macroshell (TM) capsules (large fuel containers) concurrently. The principle processes of the system are: (1) storage of the tritium as a uranium hydride; (2) pressure intensification using cryogenics; and (3) filling of the shells by permeation at elevated temperatures. The design of the facility was centered around a NRC license limit of 6000 Ci

IFMIF - International Fusion Materials Irradiation Facility Conceptual Design Activity/Interim Report

International Nuclear Information System (INIS)

Rennich, M.J.

1995-12-01

Environmental acceptability, safety, and economic viability win ultimately be the keys to the widespread introduction of fusion power. This will entail the development of radiation- resistant and low- activation materials. These low-activation materials must also survive exposure to damage from neutrons having an energy spectrum peaked near 14 MeV with annual radiation doses in the range of 20 displacements per atom (dpa). Testing of candidate materials, therefore, requires a high-flux source of high energy neutrons. The problem is that there is currently no high-flux source of neutrons in the energy range above a few MeV. The goal, is therefore, to provide an irradiation facility for use by fusion material scientists in the search for low-activation and damage-resistant materials. An accellerator-based neutron source has been established through a number of international studies and workshops' as an essential step for materials development and testing. The mission of the International Fusion Materials Irradiation Facility (IFMIF) is to provide an accelerator-based, deuterium-lithium (D-Li) neutron source to produce high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials up to about a full lifetime of anticipated use in fusion energy reactors. would also provide calibration and validation of data from fission reactor and other accelerator-based irradiation tests. It would generate material- specific activation and radiological properties data, and support the analysis of materials for use in safety, maintenance, recycling, decommissioning, and waste disposal systems

Mirror Fusion Test Facility (MFTF)

International Nuclear Information System (INIS)

Thomassen, K.I.

1978-01-01

A large, new Mirror Fusion Test Facility is under construction at LLL. Begun in FY78 it will be completed at the end of FY78 at a cost of $94.2M. This facility gives the mirror program the flexibility to explore mirror confinement principles at a signficant scale and advances the technology of large reactor-like devices. The role of MFTF in the LLL program is described here

Technology of mirror machines: LLL facilities for magnetic mirror fusion experiments

International Nuclear Information System (INIS)

Batzer, T.H.

1977-01-01

Significant progress in plasma confinement and temperature has been achieved in the 2XIIB facility at Livermore. These encouraging results, and their theoretical corroboration, have provided a firm basis for the design of a new generation of magnetic mirror experiments, adding support to the mirror concept of a fusion reactor. Two new mirror experiments have been proposed to succeed the currently operating 2XIIB facility. The first of these called TMX (Tandem Mirror Experiment) has been approved and is currently under construction. TMX is designed to utilize the intrinsic positive plasma potential of two strong, and relatively small, minimum B mirror cells to enhance the confinement of a much larger, magnetically weaker, centrally-located mirror cell. The second facility, MFTF (Mirror Fusion Test Facility), is currently in preliminary design with line item approval anticipated for FY 78. MFTF is designed primarily to exploit the experimental and theoretical results derived from 2XIIB. Beyond that, MFTF will develop the technology for the transition from the present small mirror experiments to large steady-state devices such as the mirror FERF/FTR. The sheer magnitude of the plasma volume, magnetic field, neutral beam power, and vacuum pumping capacity, particularly in the case of MFTF, has placed new and exciting demands on engineering technology. An engineering overview of MFTF, TMX, and associated MFE activities at Livermore will be presented

Fusion Engineering Device. Volume VI. Complementary development plan for engineering development

International Nuclear Information System (INIS)

1981-10-01

The basic approach followed in this volume is to define key technical issues for several fusion reactor technologies and to device program strategies to resolve each of these issues. Particular attention has been paid to elucidating the role of FED vis-a-vis complementary (non-FED) facilities in this process. The remainder of this chapter consists of summaries of the major conclusions of the technology plans in each of the areas studied, i.e., plasma heating, magnetics, nuclear, and systems considerations

Operating procedures: Fusion Experiments Analysis Facility

Energy Technology Data Exchange (ETDEWEB)

Lerche, R.A.; Carey, R.W.

1984-03-20

The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility.

Operating procedures: Fusion Experiments Analysis Facility

International Nuclear Information System (INIS)

Lerche, R.A.; Carey, R.W.

1984-01-01

The Fusion Experiments Analysis Facility (FEAF) is a computer facility based on a DEC VAX 11/780 computer. It became operational in late 1982. At that time two manuals were written to aid users and staff in their interactions with the facility. This manual is designed as a reference to assist the FEAF staff in carrying out their responsibilities. It is meant to supplement equipment and software manuals supplied by the vendors. Also this manual provides the FEAF staff with a set of consistent, written guidelines for the daily operation of the facility

Code of a Tokamak Fusion Energy Facility ITER

International Nuclear Information System (INIS)

Yasuhide Asada; Kenzo Miya; Kazuhiko Hada; Eisuke Tada

2002-01-01

The technical structural code for ITER (International Thermonuclear Experimental Fusion Reactor) and, as more generic applications, for D-T burning fusion power facilities (hereafter, Fusion Code) should be innovative because of their quite different features of safety and mechanical components from nuclear fission reactors, and the necessity of introducing several new fabrication and examination technologies. Introduction of such newly developed technologies as inspection-free automatic welding into the Fusion Code is rationalized by a pilot application of a new code concept of s ystem-based code for integrity . The code concept means an integration of element technical items necessary for construction, operation and maintenance of mechanical components of fusion power facilities into a single system to attain an optimization of the total margin of these components. Unique and innovative items of the Fusion Code are typically as follows: - Use of non-metals; - Cryogenic application; - New design margins on allowable stresses, and other new design rules; - Use of inspection-free automatic welding, and other newly developed fabrication technologies; - Graded approach of quality assurance standard to cover radiological safety-system components as well as non-safety-system components; - Consideration on replacement components. (authors)

MRI Volume Fusion Based on 3D Shearlet Decompositions.

Science.gov (United States)

Duan, Chang; Wang, Shuai; Wang, Xue Gang; Huang, Qi Hong

2014-01-01

Nowadays many MRI scans can give 3D volume data with different contrasts, but the observers may want to view various contrasts in the same 3D volume. The conventional 2D medical fusion methods can only fuse the 3D volume data layer by layer, which may lead to the loss of interframe correlative information. In this paper, a novel 3D medical volume fusion method based on 3D band limited shearlet transform (3D BLST) is proposed. And this method is evaluated upon MRI T2* and quantitative susceptibility mapping data of 4 human brains. Both the perspective impression and the quality indices indicate that the proposed method has a better performance than conventional 2D wavelet, DT CWT, and 3D wavelet, DT CWT based fusion methods.

MRI Volume Fusion Based on 3D Shearlet Decompositions

Directory of Open Access Journals (Sweden)

Chang Duan

2014-01-01

Full Text Available Nowadays many MRI scans can give 3D volume data with different contrasts, but the observers may want to view various contrasts in the same 3D volume. The conventional 2D medical fusion methods can only fuse the 3D volume data layer by layer, which may lead to the loss of interframe correlative information. In this paper, a novel 3D medical volume fusion method based on 3D band limited shearlet transform (3D BLST is proposed. And this method is evaluated upon MRI T2* and quantitative susceptibility mapping data of 4 human brains. Both the perspective impression and the quality indices indicate that the proposed method has a better performance than conventional 2D wavelet, DT CWT, and 3D wavelet, DT CWT based fusion methods.

A spallation-based irradiation test facility for fusion and future fission materials

International Nuclear Information System (INIS)

Samec, K.; Fusco, Y.; Kadi, Y.; Luis, R.; Romanets, Y.; Behzad, M.; Aleksan, R.; Bousson, S.

2014-01-01

The EU's FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the proposed DEMO fusion reactor, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550 deg. C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum over a volume occupying one litre. The entire 'TMIF' facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility. (authors)

Mirror Fusion Test Facility magnet

International Nuclear Information System (INIS)

Henning, C.H.; Hodges, A.J.; Van Sant, J.H.; Hinkle, R.E.; Horvath, J.A.; Hintz, R.E.; Dalder, E.; Baldi, R.; Tatro, R.

1979-01-01

The Mirror Fusion Test Facility (MFTF) is the largest of the mirror program experiments for magnetic fusion energy. It seeks to combine and extend the near-classical plasma confinement achieved in 2XIIB with the most advanced neutral-beam and magnet technologies. The product of ion density and confinement time will be improved more than an order of magnitude, while the superconducting magnet weight will be extrapolated from the 15 tons in Baseball II to 375 tons in MFTF. Recent reactor studies show that the MFTF will traverse much of the distance in magnet technology towards the reactor regime. Design specifics of the magnet are given

Fusion technology development: role of fusion facility upgrades and fission test reactors

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.

1983-01-01

The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. The authors show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

Fusion technology development: role of fusion facility upgrades and fission test reactors

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Miller, L.G.; Longhurst, G.R.; Schmunk, R.E.

1983-01-01

The near term national fusion program is unlikely to follow the aggressive logic of the Fusion Engineering Act of 1980. Faced with level budgets, a large, new fusion facility with an engineering thrust is unlikely in the near future. Within the fusion community the idea of upgrading the existing machines (TFTR, MFTF-B) is being considered to partially mitigate the lack of a design data base to ready the nation to launch an aggressive, mission-oriented fusion program with the goal of power production. This paper examines the cost/benefit issues of using fusion upgrades to develop the technology data base which will be required to support the design and construction of the next generation of fusion machines. The extent of usefulness of the nation's fission test reactors will be examined vis-a-vis the mission of the fusion upgrades. We will show that while fission neutrons will provide a useful test environment in terms of bulk heating and tritium breeding on a submodule scale, they can play only a supporting role in designing the integrated whole modules and systems to be used in a nuclear fusion machine

Operating large controlled thermonuclear fusion research facilities

International Nuclear Information System (INIS)

Gaudreau, M.P.J.; Tarrh, J.M.; Post, R.S.; Thomas, P.

1987-01-01

The MIT Tara Tandem Mirror is a large, state of the art controlled thermonuclear fusion research facility. Over the six years of its design, implementation, and operation, every effort was made to minimize cost and maximize performance by using the best and latest hardware, software, and scientific and operational techniques. After reviewing all major DOE fusion facilities, an independent DOE review committee concluded that the Tara operation was the most automated and efficient of all DOE facilities. This paper includes a review of the key elements of the Tara design, construction, operation, management, physics milestones, and funding that led to this success. The authors emphasize a chronological description of how the system evolved from the proposal stage to a mature device with an emphasis on the basic philosophies behind the implementation process. This description can serve both as a qualitative and quantitative database for future large experiment planning. It includes actual final costs and manpower spent as well as actual run and maintenance schedules, number of data shots, major system failures, etc. The paper concludes with recommendations for the next generation of facilities

The international fusion materials irradiation facility

International Nuclear Information System (INIS)

Shannon, T.E.; Cozzani, F.; Crandall, D.H.; Wiffen, F.W.; Katsuta, H.; Kondo, T.; Teplyakov, V.; Zavialsky, L.

1994-01-01

It is widely agreed that the development of materials for fusion systems requires a high flux, 14 MeV neutron source. The European Union, Japan, Russia and the US have initiated the conceptual design of such a facility. This activity, under the International Energy Agency (IEA) Fusion Materials Agreement, will develop the design for an accelerator-based D-Li system. The first organizational meeting was held in June 1994. This paper describes the system to be studied and the approach to be followed to complete the conceptual design by early 1997

An advanced fusion neutron source facility

International Nuclear Information System (INIS)

Smith, D.L.

1992-01-01

Accelerator-based 14-MeV-neutron sources based on modifications of the original Fusion Materials Irradiation Facility are currently under consideration for investigating the effects of high-fluence high-energy neutron irradiation on fusion-reactor materials. One such concept for a D-Li neutron source is based on recent advances in accelerator technology associated with the Continuous Wave Deuterium Demonstrator accelerator under construction at Argonne National Laboratory, associated superconducting technology, and advances in liquid-metal technology. In this paper a summary of conceptual design aspects based on improvements in technologies is presented

Remote operations in a Fusion Engineering Research Facility (FERF)

International Nuclear Information System (INIS)

Doggett, J.N.

1975-01-01

The proposed Fusion Engineering Research Facility (FERF) has been designed for the test and evaluation of materials that will be exposed to the hostile radiation environment created by fusion reactors. Because the FERF itself must create a very hostile radiation environment, extensive remote handling procedures will be required as part of its routine operations as well as for both scheduled and unscheduled maintenance. This report analyzes the remote-handling implications of a vertical- rather than horizontal-orientation of the FERF magnet, describes the specific remote-handling facilities of the proposed FERF installation and compares the FERF remote-handling system with several other existing and proposed facilities. (U.S.)

Cryogenic systems for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Slack, D.S.; Chronis, W.C.; Nelson, R.L.

1986-01-01

This paper will include an in-depth discussion of the design, fabrication, and operation of the Mirror Fusion Test Facility (MFTF) cryogenic system located at Lawrence Livermore National Laboratory (LLNL). Each subsystem will be discussed to present a basic composite of the entire facility

RTNS-II fusion materials irradiation facility

International Nuclear Information System (INIS)

Heikkinen, D.W.; Tuckerman, D.B.; Davis, J.C.; Massoletti, D.J.; Short, D.W.

1986-01-01

The Rotating Target Neutron Source (RTNS-II) facility provides an intense source of 14-MeV neutrons for the fusion energy programs of Japan and the United States. Each of the two identical accelerator-based neutron sources is capable of providing source strengths in excess of 3 x 10 13 n/s using deuteron beam currents up to 150 mA. The present status of the facility, as well as the various upgrade options, will be described in detail

Remote-handling demonstration tests for the Fusion Materials Irradiation Test (FMIT) Facility

International Nuclear Information System (INIS)

Shen, E.J.; Hussey, M.W.; Kelly, V.P.; Yount, J.A.

1982-01-01

The mission of the Fusion Materials Irradiation Test (FMIT) Facility is to create a fusion-like environment for fusion materials development. Crucial to the success of FMIT is the development and testing of remote handling systems required to handle materials specimens and maintenance of the facility. The use of full scale mock-ups for demonstration tests provides the means for proving these systems

Helios, a 20 TW CO2 laser fusion facility

International Nuclear Information System (INIS)

Ladish, J.S.

1979-01-01

Since June 1978 the Los Alamos Scientific Laboratory's Helios CO 2 laser fusion facility has been committed to an experimental target program to investigate the feasibility of laser produced inertial confinement fusion. This system is briefly described, and preliminary experimental results are reported

Local bone graft harvesting and volumes in posterolateral lumbar fusion: a technical report.

Science.gov (United States)

Carragee, Eugene J; Comer, Garet C; Smith, Micah W

2011-06-01

In lumbar surgery, local bone graft is often harvested and used in posterolateral fusion procedures. The volume of local bone graft available for posterolateral fusion has not been determined in North American patients. Some authors have described this as minimal, but others have suggested the volume was sufficient to be reliably used as a stand-alone bone graft substitute for single-level fusion. To describe the technique used and determine the volume of local bone graft available in a cohort of patients undergoing single-level primary posterolateral fusion by the authors harvesting technique. Technical description and cohort report. Consecutive patients undergoing lumbar posterolateral fusion with or without instrumentation for degenerative processes. Local bone graft volume. Consecutive patients undergoing lumbar posterolateral fusion with or without instrumentation for degenerative processes of were studied. Local bone graft was harvested by a standard method in each patient and the volume measured by a standard procedure. Twenty-five patients were studied, and of these 11 (44%) had a previous decompression. The mean volume of local bone graft harvested was measured to be 25 cc (range, 12-36 cc). Local bone graft was augmented by iliac crest bone in six of 25 patients (24%) if the posterolateral fusion bed was not well packed with local bone alone. There was a trend to greater local bone graft volumes in men and in patients without previous decompression. Large volumes of local bone can be harvested during posterolateral lumbar fusion surgery. Even in patients with previous decompression the volume harvested is similar to that reported harvested from the posterior iliac crest for single-level fusion. Copyright © 2011 Elsevier Inc. All rights reserved.

Volume ignition of laser driven fusion pellets and double layer effects

International Nuclear Information System (INIS)

Cicchitelli, L.; Eliezer, S.; Goldsworthy, M.P.; Green, F.; Hora, H.; Ray, P.S.; Stening, R.J.; Szichman, H.

1988-01-01

The realization of an ideal volume compression of laser-irradiated fusion pellets opens the possibility for an alternative to spark ignition proposed for many years for inertial confinement fusion. A re-evaluation of the difficulties of the central spark ignition of laser driven pellets is given. The alternative volume compression theory, together with volume burn and volume ignition, have received less attention and are re-evaluated in view of the experimental verification generalized fusion gain formulas, and the variation of optimum temperatures derived at self-ignition. Reactor-level DT fusion with MJ-laser pulses and volume compression to 50 times the solid-state density are estimated. Dynamic electric fields and double layers at the surface and in the interior of plasmas result in new phenomena for the acceleration of thermal electrons to suprathermal electrons. Double layers also cause a surface tension which stabilizes against surface wave effects and Rayleigh-Taylor instabilities. (author)

Performance of Shiva as a laser fusion irradiation facility

International Nuclear Information System (INIS)

Speck, D.R.; Bliss, E.S.; Glaze, J.A.; Johnson, B.C.; Manes, K.R.; Ozarski, R.G.; Rupert, P.R.; Simmons, W.W.; Swift, C.D.; Thompson, C.E.

1979-01-01

Shiva is a 20 beam Nd:Glass Laser and Target Irradiation Facility at the Lawrence Livermore Laboratory. The laser system and integrated target facility evolved during the last year from a large, untested, experimental laser system to a target irradiation facility which has provided significant laser driven inertial confinement fusion data. The operation of the facility is discussed

Operating large controlled thermonuclear fusion research facilities

International Nuclear Information System (INIS)

Gaudreau, M.P.J.; Tarrh, J.M.; Post, R.S.; Thomas, P.

1987-10-01

The MIT Tara Tandem Mirror is a large, state of the art controlled thermonuclear fusion research facility. Over the six years of its design, implementation, and operation, every effort was made to minimize cost and maximize performance by using the best and latest hardware, software, and scientific and operational techniques. After reviewing all major DOE fusion facilities, an independent DOE review committee concluded that the Tara operation was the most automated and efficient of all DOE facilities. This paper includes a review of the key elements of the Tara design, construction, operation, management, physics milestones, and funding that led to this success. We emphasize a chronological description of how the system evolved from the proposal stage to a mature device with an emphasis on the basic philosophies behind the implementation process. This description can serve both as a qualitative and quantitative database for future large experiment planning. It includes actual final costs and manpower spent as well as actual run and maintenance schedules, number of data shots, major system failures, etc. The paper concludes with recommendations for the next generation of facilities. 13 refs., 15 figs., 3 tabs

The FENIX [Fusion ENgineering International EXperimental] test facility

International Nuclear Information System (INIS)

Slack, D.S.; Patrick, R.E.; Chaplin, M.R.; Miller, J.R.; Shen, S.S.; Summers, L.T.; Kerns, J.A.

1989-01-01

The Fusion ENgineering International EXperimental Magnet Facility (FENIX), under construction at Lawrence Livermore National Laboratory (LLNL), is a significant step forward in meeting the testing requirements necessary for the development of superconductor for large-scale, superconducting magnets. A 14-T, transverse field over a test volume of 150 x 60 x 150 mm in length will be capable of testing conductors the size of the International Thermonuclear Experimental Reactor (ITER). Proposed conductors for ITER measure ∼35 mm on one side and will operate at currents of up to 40 kA at fields of ∼14 T. The testing of conductors and associated components, such as joints, will require large-bore, high-field magnet facilities. FENIX is being constructed using the existing A 2o and A 2i magnets from the idle MFTF. The east and west A 2 pairs will be mounted together to form a split-pair solenoid. The pairs of magnets will be installed in a 4.0-m cryostat vessel located in the HFTF building at LLNL. Each magnet is enclosed in its own cryostat, the existing 4.0-m vessel serving only as a vacuum chamber. 4 refs., 8 figs

Radiological design criteria for fusion power test facilities

International Nuclear Information System (INIS)

Singh, M.S.; Campbell, G.W.

1982-01-01

The quest for fusion power and understanding of plasma physics has resulted in planning, design, and construction of several major fusion power test facilities, based largely on magnetic and inertial confinement concepts. We have considered radiological design aspects of the Joint European Torus (JET), Livermore Mirror and Inertial Fusion projects, and Princeton Tokamak. Our analyses on radiological design criteria cover acceptable exposure levels at the site boundary, man-rem doses for plant personnel and population at large, based upon experience gained for the fission reactors, and on considerations of cost-benefit analyses

Identification and selection of initiating events for experimental fusion facilities

International Nuclear Information System (INIS)

Cadwallader, L.C.

1989-01-01

This paper describes the current approaches used in probabilistic risk assessment (PRA) to identify and select accident initiating events for study in either probabilistic safety analysis or PRA. Current methods directly apply to fusion facilities as well as other types of industries, such as chemical processing and nuclear fission. These identification and selection methods include the Master Logic Diagram, historical document review, system level Failure Modes and Effects Analysis, and others. A combination of the historical document review, such as Safety Analysis Reports and fusion safety studies, and the Master Logic Diagram with appropriate quality assurance reviews, is suggested for standardizing US fusion PRA effects. A preliminary set of generalized initiating events applicable to fusion facilities derived from safety document review is presented as a framework to start from for the historical document review and Master Logic Diagram approach. Fusion designers should find this list useful for their design reviews. 29 refs., 2 tabs

Identification and selection of initiating events for experimental fusion facilities

International Nuclear Information System (INIS)

Cadwallader, L.C.

1989-01-01

This paper describes the current approaches used in probabilistic risk assessment (PRA) to identify and select accident initiating events for study in either probabilistic safety analysis or PRA. Current methods directly apply to fusion facilities as well as other types of industries, such as chemical processing and nuclear fission. These identification and selection methods include the Master Logic Diagram, historical document review, system level Failure Modes and Effects Analysis, and others. A combination of the historical document review, such as Safety Analysis Reports and fusion safety studies, and the Master Logic Diagram with appropriate quality assurance reviews, is suggested for standardizing U.S. fusion PRA efforts. A preliminary set of generalized initiating events applicable to fusion facilities derived from safety document review is presented as a framework to start from for the historical document review and Master Logic Diagram approach. Fusion designers should find this list useful for their design reviews. 29 refs., 1 tab

The National Ignition Facility (NIF): A path to fusion energy

International Nuclear Information System (INIS)

Moses, Edward I.

2008-01-01

Fusion energy has long been considered a promising, clean, nearly inexhaustible source of energy. Power production by fusion micro-explosions of inertial confinement fusion (ICF) targets has been a long-term research goal since the invention of the first laser in 1960. The National Ignition Facility (NIF) is poised to take the next important step in the journey by beginning experiments researching ICF ignition. Ignition on NIF will be the culmination of over 30 years of ICF research on high-powered laser systems such as the Nova laser at Lawrence Livermore National Laboratory (LLNL) and the OMEGA laser at the University of Rochester, as well as smaller systems around the world. NIF is a 192-beam Nd-glass laser facility at LLNL that is more than 90% complete. The first cluster of 48 beams is operational in the laser bay, the second cluster is now being commissioned, and the beam path to the target chamber is being installed. The Project will be completed in 2009, and ignition experiments will start in 2010. When completed, NIF will produce up to 1.8 MJ of 0.35-μm light in highly shaped pulses required for ignition. It will have beam stability and control to higher precision than any other laser fusion facility. Experiments using one of the beams of NIF have demonstrated that NIF can meet its beam performance goals. The National Ignition Campaign (NIC) has been established to manage the ignition effort on NIF. NIC has all of the research and development required to execute the ignition plan and to develop NIF into a fully operational facility. NIF will explore the ignition space, including direct drive, 2ω ignition, and fast ignition, to optimize target efficiency for developing fusion as an energy source. In addition to efficient target performance, fusion energy requires significant advances in high-repetition-rate lasers and fusion reactor technology. The Mercury laser at LLNL is a high-repetition-rate Nd-glass laser for fusion energy driver development. Mercury

Trends of researches for fusion engineering research facility (FERF)

International Nuclear Information System (INIS)

Ozawa, Yasutomo; Enoto, Takeaki

1975-01-01

The role of a fusion neutron radiation test facility in the development of a scientific feasibility experimental reactor or demonstration fusion power reactor plant would be analogous to the role of the materials testing and experimental reactors in the development of fission power reactor. While the material testing fission reactor has been developed after successful operation of fission reactors, in the case of fusion reactor development it is desirable to realize the fusion engineering research facility (FERF) in-phase to the development of SFX and/or demonstration fusion power reactor plants. Here so called FERF in near future is the Controlled Thermonuclear Reactor which provides the high-intensity and high-energy neutron and plasma source whether the net power output is produced or not. From the point of direct attainment to SFX, we would like to emphasize that FEFE is the royal road leading to the goal of successful achievement of CTR program and could be useful for the experiment on impurity effects caused by neutron and plasma irradiations onto the wall material for SFX. Further, we rather suppose that hybrid FERF-fission assembly could be fairly and easily realizable in near future. (auth.)

Neutronics analysis of International Fusion Material Irradiation Facility (IFMIF). Japanese contributions

International Nuclear Information System (INIS)

Oyama, Yukio; Noda, Kenji; Kosako, Kazuaki.

1997-10-01

In fusion reactor development for demonstration reactor, i.e., DEMO, materials tolerable for D-T neutron irradiation are absolutely required for both mechanical and safety point of views. For this requirement, several kinds of low activation materials were proposed. However, experimental data by actual D-T fusion neutron irradiation have not existed so far because of lack of fusion neutron irradiation facility, except fundamental radiation damage studies at very low neutron fluence. Therefore such a facility has been strongly requested. According to agreement of need for such a facility among the international parties, a conceptual design activity (CDA) of International Fusion Material Irradiation Facility (IFMIF) has been carried out under the frame work of the IEA-Implementing Agreement. In the activity, a neutronics analysis on irradiation field optimization in the IFMIF test cell was performed in three parties, Japan, US and EU. As the Japanese contribution, the present paper describes a neutron source term as well as incident deuteron beam angle optimization of two beam geometry, beam shape (foot print) optimization, and dpa, gas production and heating estimation inside various material loading Module, including a sensitivity analysis of source term uncertainty to the estimated irradiation parameters. (author)

Simulation for evaluation of the multi-ion-irradiation Laboratory of TechnoFusion facility and its relevance for fusion applications

International Nuclear Information System (INIS)

Jimenez-Rey, D.; Mota, F.; Vila, R.; Ibarra, A.; Ortiz, Christophe J.; Martinez-Albertos, J.L.; Roman, R.; Gonzalez, M.; Garcia-Cortes, I.; Perlado, J.M.

2011-01-01

Thermonuclear fusion requires the development of several research facilities, in addition to ITER, needed to advance the technologies for future fusion reactors. TechnoFusion will focus in some of the priority areas identified by international fusion programmes. Specifically, the TechnoFusion Area of Irradiation of Materials aims at surrogating experimentally the effects of neutron irradiation on materials using a combination of ion beams. This paper justifies this approach using computer simulations to validate the multi-ion-irradiation Laboratory. The planned irradiation facility will investigate the effects of high energetic radiations on reactor-relevant materials. In a second stage, it will also be used to analyze the performance of such materials and evaluate newly designed materials. The multi-ion-irradiation Laboratory, both triple irradiation and high-energy proton irradiation, can provide valid experimental techniques to reproduce the effect of neutron damage in fusion environment.

[Glass Development Laser (GDL) Facility upgrade.] LLE Review. Quarterly report, October-December 1984. Volume 21

International Nuclear Information System (INIS)

Kim, H.

1984-01-01

This volume of the LLE Review contains articles on the upgrade of the GDL (Glass Development) system, theoretical advances in the laser fusion effort, improved target fabrication capabilities, x-ray laser research, developments in the picosecond optics research of the LLE advanced technology program, and on the National Laser Users Facility activities for October-December 1984. 56 refs., 31 figs., 3 tabs

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

International Nuclear Information System (INIS)

Mota, F.; Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V.

2011-01-01

Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al 2 O 3 , SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

Analysis of displacement damage in materials in nuclear fusion facilities (DEMO, IFMIF and TechnoFusion)

Energy Technology Data Exchange (ETDEWEB)

Mota, F., E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain); Vila, R.; Ortiz, C.; Garcia, A.; Casal, N.; Ibarra, A.; Rapisarda, D.; Queral, V. [Laboratorio Nacional de Fusion por Confinamiento Magnetico-CIEMAT, 28040 Madrid (Spain)

2011-10-15

Present pathway to fusion reactors includes a rigorous material testing program. To reach this objective, irradiation facilities must produce the displacement damage per atom (dpa), primary knock-on atom (PKA) spectrum and gaseous elements by transmutation reactions (He, H) as closely as possible to the ones expected in the future fusion reactors (as DEMO).The irradiation parameters (PKA spectra and damage function) of some candidate materials for fusion reactors (Al{sub 2}O{sub 3}, SiC and Fe) have been studied and then, the suitability of some proposed experimental facilities, such as IFMIF and TechnoFusion, to perform relevant tests with these materials has been assessed.The following method has been applied: neutron fluxes present in different irradiation modules of IFMIF have been calculated by the neutron transport McDeLicious code. In parallel, the energy differential cross sections of PKA have been calculated by using the NJOY code. After that, the damage generated by the PKA spectra was analyzed using the MARLOWE code (binary collision approximation) and custom analysis codes. Finally, to analyze the ions effects in different irradiation conditions in the TechnoFusion irradiation area, the SRIM and Marlowe codes have been used. The results have been compared with the expected ones for a DEMO HCLL reactor.

Fusion Technology 1996. Proceedings. Volume 1 and 2

International Nuclear Information System (INIS)

Varandas, C.; Serra, F.

1997-01-01

The objective of these proceedings was to provide a platform for the exchange of information on the design, construction and operation of fusion experiments. The technology which is being developed for the next step devices and fusion reactors was also covered. Sections in volume 1 concern (A) first wall, divertors and vacuum systems; (B) plasma heating and control; (C) plasma engineering and control; and (D) experimental systems. The sections in volume 2 deal with (E) magnet and related power supplies; (F) fuel cycle and tritium processing systems; (G) blanket technology/materials; (H) assembly, remote handling and waste management and storage; and (I) safety and environment, and reactor studies

Integrated operations plan for the MFTF-B Mirror Fusion Test Facility. Volume II. Integrated operations plan

Energy Technology Data Exchange (ETDEWEB)

1981-12-01

This document defines an integrated plan for the operation of the Lawrence Livermore National Laboratory (LLNL) Mirror Fusion Test Facility (MFTF-B). The plan fulfills and further delineates LLNL policies and provides for accomplishing the functions required by the program. This plan specifies the management, operations, maintenance, and engineering support responsibilities. It covers phasing into sustained operations as well as the sustained operations themselves. Administrative and Plant Engineering support, which are now being performed satisfactorily, are not part of this plan unless there are unique needs.

Integrated operations plan for the MFTF-B Mirror Fusion Test Facility. Volume II. Integrated operations plan

International Nuclear Information System (INIS)

1981-12-01

This document defines an integrated plan for the operation of the Lawrence Livermore National Laboratory (LLNL) Mirror Fusion Test Facility (MFTF-B). The plan fulfills and further delineates LLNL policies and provides for accomplishing the functions required by the program. This plan specifies the management, operations, maintenance, and engineering support responsibilities. It covers phasing into sustained operations as well as the sustained operations themselves. Administrative and Plant Engineering support, which are now being performed satisfactorily, are not part of this plan unless there are unique needs

Compressed Gas Safety for Experimental Fusion Facilities

Energy Technology Data Exchange (ETDEWEB)

Lee C. Cadwallader

2004-09-01

Experimental fusion facilities present a variety of hazards to the operators and staff. There are unique or specialized hazards, including magnetic fields, cryogens, radio frequency emissions, and vacuum reservoirs. There are also more general industrial hazards, such as a wide variety of electrical power, pressurized air, and cooling water systems in use, there are crane and hoist loads, working at height, and handling compressed gas cylinders. This paper outlines the projectile hazard assoicated with compressed gas cylinders and mthods of treatment to provide for compressed gas safety. This information should be of interest to personnel at both magnetic and inertial fusion experiments.

Neutron-irradiation facilities at the Intense Pulsed Neutron Source-I for fusion magnet materials studies

International Nuclear Information System (INIS)

Brown, B.S.; Blewitt, T.H.

1982-01-01

The decommissioning of reactor-based neutron sources in the USA has led to the development of a new generation of neutron sources that employ high-energy accelerators. Among the accelerator-based neutron sources presently in operation, the highest-flux source is the Intense Pulsed Neutron Source (IPNS), a user facility at Argonne National Laboratory. Neutrons in this source are produced by the interaction of 400 to 500 MeV protons with either of two 238 U target systems. In the Radiation Effects Facility (REF), the 238 U target is surrounded by Pb for neutron generatjion and reflection. The REF has three separate irradiation thimbles. Two thimbles provide irradiation temperatures between that of liquid He and several hundred degrees centigrade. The third thimble operates at ambient temperature. The large irradiation volume, the neutron spectrum and flux, the ability to transfer samples without warm up, and the dedication of the facilities during the irradiation make this ideally suited for radiation damage studies on components for superconducting fusion magnets. Possible experiments for fusion magnet materials are discussed on cyclic irradiation and annealing of stabilizers in a high magnetic field, mechanical tests on organic insulation irradiated at 4 K, and superconductors measured in high fields after irradiation

Experimental fusion power reactor conceptual design study. Final report. Volume II

International Nuclear Information System (INIS)

Baker, C.C.

1976-12-01

This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following sections: (1) reactor components, (2) auxiliary systems, (3) operations, (4) facility design, (5) program considerations, and (6) conclusions and recommendations

Conceptual design of a mirror reactor for a fusion engineering research facility (FERF)

International Nuclear Information System (INIS)

Batzer, T.H.; Burleigh, R.C.; Carlson, G.A.; Dexter, W.L.; Hamilton, G.W.; Harvey, A.R.; Hickman, R.G.; Hoffman, M.A.; Hooper, E.B. Jr.; Moir, R.W.; Nelson, R.L.; Pittenger, L.C.; Smith, B.H.; Taylor, C.E.; Werner, R.W.; Wilcox, T.P.

1975-01-01

A conceptual design is presented for a small mirror fusion reactor for a Fusion Engineering Research Facility (FERF). The reactor produces 3.4 MW of fusion power and a useful neutron flux of about 10 14 n.cm -2 .s -1 . Superconducting ''yin-yang'' coils are used, and the plasma is sustained by injection of energetic neutral D 0 and T 0 . Conceptual layouts are given for the reactor, its major components, and supporting facilities. (author)

Mirror fusion test facility

International Nuclear Information System (INIS)

Post, R.F.

1978-01-01

The MFTF is a large new mirror facility under construction at Livermore for completion in 1981--82. It represents a scaleup, by a factor of 50 in plasma volume, a factor of 5 or more in ion energy, and a factor of 4 in magnetic field intensity over the Livermore 2XIIB experiment. Its magnet, employing superconducting NbTi windings, is of Yin-Yang form and will weigh 200 tons. MFTF will be driven by neutral beams of two levels of current and energy: 1000 amperes of 20 keV (accelerating potential) pulsed beams for plasma startup; 750 amperes of 80 keV beams of 0.5 second duration for temperature buildup and plasma sustainment. Two operating modes for MFTF are envisaged: The first is operation as a conventional mirror cell with n/sup tau/ approximately equal to 10 12 cm -3 sec, W/sub i/ = 50 keV, where the emphasis will be on studying the physics of mirror cells, particularly the issues of improved techniques of stabilization against ion cyclotron modes and of maximization of the electron temperature. The second possible mode is the further study of the Field Reversed Mirror idea, using high current neutral beams to sustain the field-reversed state. Anticipating success in the coming Livermore Tandem Mirror Experiment (TMX) MFTF has been oriented so that it could comprise one end cell of a scaled up TM experiment. Also, if MFTF were to succeed in achieving a FR state it could serve as an essentially full-sized physics prototype of one cell of a FRM fusion power plant

Tandem mirror magnet system for the mirror fusion test facility

International Nuclear Information System (INIS)

Bulmer, R.H.; Van Sant, J.H.

1980-01-01

The Tandem Mirror Fusion Test Facility (MFTF-B) will be a large magnetic fusion experimental facility containing 22 supercounducting magnets including solenoids and C-coils. State-of-the-art technology will be used extensively to complete this facility before 1985. Niobium titanium superconductor and stainless steel structural cases will be the principle materials of construction. Cooling will be pool boiling and thermosiphon flow of 4.5 K liquid helium. Combined weight of the magnets will be over 1500 tonnes and the stored energy will be over 1600 MJ. Magnetic field strength in some coils will be more than 8 T. Detail design of the magnet system will begin early 1981. Basic requirements and conceptual design are disclosed in this paper

Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

International Nuclear Information System (INIS)

Beller, D.E.; Ott, K.O.; Terry, W.K.

1987-01-01

A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

The national ignition facility (NIF) : A path to fusion energy

International Nuclear Information System (INIS)

Moses, E. I.

2007-01-01

Fusion energy has long been considered a promising clean, nearly inexhaustible source of energy. Power production by fusion micro-explosions of inertial confinement fusion (ICF) targets has been a long term research goal since the invention of the first laser in 1960. The NIF is poised to take the next important step in the journey by beginning experiments researching ICF ignition. Ignition on NIF will be the culmination of over thirty years of ICF research on high-powered laser systems such as the Nova laser at LLNL and the OMEGA laser at the University of Rochester as well as smaller systems around the world. NIF is a 192 beam Nd-glass laser facility at LLNL that is more than 90% complete. The first cluster of 48 beams is operational in the laser bay, the second cluster is now being commissioned, and the beam path to the target chamber is being installed. The Project will be completed in 2009 and ignition experiments will start in 2010. When completed NIF will produce up to 1.8 MJ of 0.35 μm light in highly shaped pulses required for ignition. It will have beam stability and control to higher precision than any other laser fusion facility. Experiments using one of the beams of NIF have demonstrated that NIF can meet its beam performance goals. The National Ignition Campaign (NIC) has been established to manage the ignition effort on NIF. NIC has all of the research and development required to execute the ignition plan and to develop NIF into a fully operational facility. NIF will explore the ignition space, including direct drive, 2ω ignition, and fast ignition, to optimize target efficiency for developing fusion as an energy source. In addition to efficient target performance, fusion energy requires significant advances in high repetition rate lasers and fusion reactor technology. The Mercury laser at LLNL is a high repetition rate Nd-glass laser for fusion energy driver development. Mercury uses state-o-the art technology such as ceramic laser slabs and light

Higher-genus su(N) fusion multiplicities as polytope volumes

International Nuclear Information System (INIS)

Flynn, G; Rasmussen, J; Tahic, M; Walton, M A

2002-01-01

We show how higher-genus su(N) fusion multiplicities may be computed as the discretized volumes of certain polytopes. The method is illustrated by explicit analyses of some su(3) and su(4) fusions, but applies to all higher-point and higher-genus su(N) fusions. It is based on an extension of the realm of Berenstein-Zelevinsky triangles by including so-called gluing and loop-gluing diagrams. The identification of the loop-gluing diagrams is our main new result, since they enable us to characterize higher-genus fusions in terms of polytopes. Also, the genus-2 0-point su(3) fusion multiplicity is found to be a simple binomial coefficient in the affine level

Efficacy of different bone volume expanders for augmenting lumbar fusions.

Science.gov (United States)

Epstein, Nancy E

2008-01-01

A wide variety of bone volume expanders are being used in performing posterolateral lumbar noninstrumented and instrumented lumbar fusions. This article presents a review of their efficacy based on fusion rates, complications, and outcomes. Lumbar noninstrumented and instrumented fusions frequently use laminar autografts and different bone graft expanders. This review presents the utility of multiple forms/ratios of DBMs containing allografts. It also discusses the efficacy of artificial bone graft substitutes, including HA and B-TCP. Dynamic x-ray and/or CT examinations were used to document fusion in most series. Outcomes were variously assessed using Odom's criteria or different outcome questionnaires (Oswestry Questionnaire, SF-36, Dallas Pain Questionnaire, and/or Low Back Pain Rating Scale). Performing noninstrumented and instrumented lumbar posterolateral fusions resulted in comparable fusion rates in many series. Similar outcomes were also documented based on Odom's criteria or the multiple patient-based questionnaires. However, in some studies, the addition of spinal instrumentation increased the reoperation rate, operative time, blood loss, and cost. Various forms of DBMs, applied in different ratios to autografts, effectively supplemented spinal fusions in animal models and patient series. beta-Tricalcium phosphate, which is used to augment autograft fusions addressing idiopathic scoliosis or lumbar disease, also proved to be effective. Different types of bone volume expanders, including various forms of allograft-based DBMs, and artificial bone graft substitutes (HA and B-TCP) effectively promote posterolateral lumbar noninstrumented and instrumented fusions when added to autografts.

Fusion Engineering Device. Volume 1. Mission and program summary

International Nuclear Information System (INIS)

1981-10-01

This volume presents, in summary form, a recommended approach to implementing the Magnetic Fusion Energy Engineering Act of 1980. These recommendations constitute the findings of the FED Technical Management Board (TMB). The TMB and the affiliated technical managers gave particular scrutiny to elucidating the role of FED in fusion development and to defining the device mission

Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

International Nuclear Information System (INIS)

Karpenko, V.N.

1983-01-01

The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 10 14 cm -3 .s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

Cryogenic systems for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Slack, D.S.; Nelson, R.L.; Chronis, W.C.

1985-08-01

This paper includes an in-depth discussion of the design, fabrication, and operation of the Mirror Fusion Test Facility (MFTF) cryogenic system located at Lawrence Livermore National Laboratory (LLNL). Each subsystem discussed to present a basic composite of the entire facility. The following subsystems are included: 500kW nitrogen reliquefier, subcoolers, and distribution system; 15kW helium refrigerator/liquefier and distribution system; helium recovery and storage system; rough vacuum and high vacuum systems

Neutron irradiation facilities for fission and fusion reactor materials studies

International Nuclear Information System (INIS)

Rowcliffe, A.F.

1985-01-01

The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

Conceptual design of a fission-based integrated test facility for fusion reactor components

International Nuclear Information System (INIS)

Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

1982-01-01

The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

Design issues for a laboratory high gain fusion facility

International Nuclear Information System (INIS)

Hogan, W.J.

1987-01-01

In an inertial fusion laboratory high gain facility, experiments will be carried out with up to 1000 MJ of thermonuclear yield. The experiment area of such a facility will include many systems and structures that will have to operate successfully in the difficult environment created by the sudden large energy release. This paper estimates many of the nuclear effects that will occur, discusses the implied design issues and suggests possible solutions so that a useful experimental facility can be built. 4 figs

Mirror fusion test facility plasma diagnostics system

International Nuclear Information System (INIS)

Thomas, S.R. Jr.; Coffield, F.E.; Davis, G.E.; Felker, B.

1979-01-01

During the past 25 years, experiments with several magnetic mirror machines were performed as part of the Magnetic Fusion Energy (MFE) Program at LLL. The latest MFE experiment, the Mirror Fusion Test Facility (MFTF), builds on the advances of earlier machines in initiating, stabilizing, heating, and sustaining plasmas formed with deuterium. The goals of this machine are to increase ion and electron temperatures and show a corresponding increase in containment time, to test theoretical scaling laws of plasma instabilities with increased physical dimensions, and to sustain high-beta plasmas for times that are long compared to the energy containment time. This paper describes the diagnostic system being developed to characterize these plasma parameters

Inertial Confinement Fusion quarterly report, January--March 1995. Volume 5, No. 2

International Nuclear Information System (INIS)

1995-01-01

The ICF quarterly report is published by the Inertial Confinement Fusion Program at the Lawrence Livermore National Laboratory. Topics included this quarter include: the role of the National Ignition Facility in the development of Inertial Confinement Fusion, laser-plasma interactions in large gas-filled hohlraums, evolution of solid-state induction modulators for a heavy-ion recirculator, the National Ignition Facility project, and terminal-level relaxation in Nd-doped laser material

Effect of electromagnetic field in fusion facility on electronic personal dosimeter

International Nuclear Information System (INIS)

Yamada, Junya; Kawano, Takao; Uda, Tatsuhiko; Shimo, Michikuni

2010-01-01

The effect of electromagnetic field on electronic personal dosimeters in a nuclear fusion facility was examined in a Magnetic Resonance Imaging (MRI) examination room instead of a nuclear fusion facility. Three types of electronic personal dosimeters, the PDM-111, the 112, and the 117, were used as typical ones. We surveyed the electromagnetic field distribution and dosimeters were placed at locations with various strengths of the electromagnetic field. The natural radiation dose was measured for about one week. We found that while dosimeters were not affected by the electric field, they were affected by the magnetic one. Dosimeters detected radiation levels less sensitively as the magnetic field strength was increased up to 150 mT. The dosimeters underestimated the environmental radiation dose rates by about 10-30% when the magnetic field strength was larger than 150 mT. We assumed that hall-effect caused the reduction in radiation sensitivity. We concluded that the strength of the magnetic field needs to be carefully considered when an electronic personal dosimeter is used for monitoring both personal and area dose in a nuclear fusion facility. (author)

Engineering design of the Nova Laser Facility for inertial-confinement fusion

International Nuclear Information System (INIS)

Simmons, W.W.; Godwin, R.O.; Hurley, C.A.

1982-01-01

The design of the Nova Laser Facility for inertial confinement fusion experiments at Lawrence Livermore National Laboratory is presented from an engineering perspective. Emphasis is placed upon design-to-performance requirements as they impact the various subsystems that comprise this complex experimental facility

Inertial confinement fusion target insertion concepts for the National Ignition Facility

International Nuclear Information System (INIS)

Laughon, G.J.; Schultz, K.R.

1996-01-01

The National Ignition Facility (NIF) will be used to demonstrate fusion ignition in a laboratory environment in order to support development of inertial fusion as a potential fusion energy source for civilian use. However, target insertion must first be addressed before inertial fusion can become a practical energy source. Since target insertion systems currently utilized are not suitable for multiple shots in quick succession, insertion concepts involving free-falling and artificially accelerated targets are developed and evaluated against a set of predetermined guidelines. It is shown that a system involving a fast retraction positioner would be suitable. 5 refs., 4 figs

Prospects for fusion applications of reversed-field pinches

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.; Hagenson, R.L.

1985-01-01

The applicability of the Reversed-Field Pinch (RFP) as a source of fusion neutrons for use in developing key fusion nuclear technologies is examined. This Fusion Test Facility (FTF) would emphasize high neutron wall loading, small plasma volume, low fusion and driver powers, and steady-state operation. Both parametric tradeoffs based on present-day physics understanding and a conceptual design based on an approx.1-MW/m 2 (neutron) driven operation are reported. 10 refs

Remote control of a fusion facility

Energy Technology Data Exchange (ETDEWEB)

Schissel, D.P. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)], E-mail: schissel@fusion.gat.com; Abla, G.; Humphreys, D.A.; Penaflor, B.G.; Sammuli, B.S.; Walker, M.L. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States)

2009-06-15

Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaboration between experimental sites and laboratories worldwide. This scientific collaboration activity is strong at existing experimental sites, is a major element of machines just coming on line, and is also a thrust of experiments that will come on line in the next decade. Computer science research into enhancing the ability to scientifically participate in a fusion experiment remotely has been growing in size in an attempt to better address the needs of fusion scientists worldwide. The natural progression of this research is to examine how to move from remote scientific participation to remote hardware control. This paper examines the challenges associated with remote experimental device control and proposes a solution based on a semantic approach that defines a Gatekeeper software system that will be the only channel of interaction for incoming requests to the experimental site. The role of the Gatekeeper is to validate the identification and access privilege of the requestor and to ensure the validity of the proposed request. The Gatekeeper will be a modular system, transparent to end-users, and allow a high volume of activity.

Remote control of a fusion facility

International Nuclear Information System (INIS)

Schissel, D.P.; Abla, G.; Humphreys, D.A.; Penaflor, B.G.; Sammuli, B.S.; Walker, M.L.

2009-01-01

Magnetic fusion experiments keep growing in size and complexity resulting in a concurrent growth in collaboration between experimental sites and laboratories worldwide. This scientific collaboration activity is strong at existing experimental sites, is a major element of machines just coming on line, and is also a thrust of experiments that will come on line in the next decade. Computer science research into enhancing the ability to scientifically participate in a fusion experiment remotely has been growing in size in an attempt to better address the needs of fusion scientists worldwide. The natural progression of this research is to examine how to move from remote scientific participation to remote hardware control. This paper examines the challenges associated with remote experimental device control and proposes a solution based on a semantic approach that defines a Gatekeeper software system that will be the only channel of interaction for incoming requests to the experimental site. The role of the Gatekeeper is to validate the identification and access privilege of the requestor and to ensure the validity of the proposed request. The Gatekeeper will be a modular system, transparent to end-users, and allow a high volume of activity.

Status of the US inertial fusion program and the National Ignition Facility

International Nuclear Information System (INIS)

Crandall, D.H.

1997-01-01

Research programs supported by the United States Office of Inertial Fusion and the NIF are summarized. The US inertial fusion program has developed an approach to high energy density physics and fusion ignition in the laboratory relying on the current physics basis of capsule drive by lasers and on the National Ignition Facility which is under construction. (AIP) copyright 1997 American Institute of Physics

Bulk-shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

Carter, L.L.; Mann, F.M.; Morford, R.J.; Johnson, D.L.; Huang, S.T.

1982-07-01

The accelerator-based Fusion Materials Irradiation Test (FMIT) facility will provide a high-fluence, fusion-like radiation environment for the testing of materials. While the neutron spectrum produced in the forward direction by the 35 MeV deuterons incident upon a flowing lithium target is characterized by a broad peak around 14 MeV, a high energy tail extends up to about 50 MeV. Some shield design considerations are reviewed

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

International Nuclear Information System (INIS)

Martone, M.

1997-01-01

This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Final report

Energy Technology Data Exchange (ETDEWEB)

Martone, M [ENEA, Centro Ricerche Frascati, Rome (Italy)

1997-01-01

This report documents the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

Tritium burning in inertial electrostatic confinement fusion facility

Energy Technology Data Exchange (ETDEWEB)

Ohnishi, Masami, E-mail: onishi@kansai-u.ac.jp [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Yamamoto, Yasushi; Osawa, Hodaka [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Hatano, Yuji; Torikai, Yuji [Hydrogen Isotope Science Center, University of Toyama, Gofuku, Toyama 930-8555 (Japan); Murata, Isao [Faculty of Engineering Environment and Energy Department, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan); Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki [Department of Science and Engineering, Kansai University, 3-3-35 Yamate-cho, Suita, Osaka 564-8680 (Japan); Masuda, Kai [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hotta, Eiki [Interdisciplinary Graduate School of Science and Engineering, Tokyo Institute of Technology, 4259 Nagatsuda-cho, Midori-ku, Yokohama 226-8503 (Japan)

2016-11-01

Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10{sup 8}(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

Tritium burning in inertial electrostatic confinement fusion facility

International Nuclear Information System (INIS)

Ohnishi, Masami; Yamamoto, Yasushi; Osawa, Hodaka; Hatano, Yuji; Torikai, Yuji; Murata, Isao; Kamakura, Keita; Onishi, Masaaki; Miyamoto, Keiji; Konda, Hiroki; Masuda, Kai; Hotta, Eiki

2016-01-01

Highlights: • An experiment on tritium burning is conducted in an inertial electrostatic confinement fusion (IECF) facility. • A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used. • The neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. • The neutron production rate of the D–T gas mixture in 1:1 ratio is expected to be more than 10"8(1/sec) in the present D–T experiment. - Abstract: An experiment on tritium burning is conducted to investigate the enhancement in the neutron production rate in an inertial electrostatic confinement fusion (IECF) facility. The facility is designed such that it is shielded from the outside for safety against tritium and a getter pump is used for evacuating the vacuum chamber and feeding the fuel gas. A deuterium–tritium gas mixture with 93% deuterium and 7% tritium is used, and its neutron production rate is measured to be 5–8 times more than that of pure deuterium gas. Moreover, the results show good agreement with those of a simplified theoretical estimation of the neutron production rate. After tritium burning, the exhausted fuel gas undergoes a tritium recovery procedure through a water bubbler device. The amount of gaseous tritium released by the developed IECF facility after tritium burning is verified to be much less than the threshold set by regulations.

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

International Nuclear Information System (INIS)

1997-01-01

This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member

IFMIF : International Fusion Materials Irradiation Facility Conceptual Design Activity: Executive summary

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-01-01

This report is a summary of the results of the Conceptual Design Activity (CDA) on the International Fusion Materials Irradiation Facility (IFMIF), conducted during 1995 and 1996. The activity is under the auspices of the International Energy Agency (IEA) Implementing Agreement for a Programme of Research and Development on Fusion Materials. An IEA Fusion Materials Executive Subcommittee was charged with overseeing the IFMIF-CDA work. Participants in the CDA are the European Union, Japan, and the United States, with the Russian Federation as an associate member.

Safety issues relating to the design of fusion power facilities

International Nuclear Information System (INIS)

Stasko, R.R.; Wong, K.Y.; Russell, S.B.

1986-06-01

In order to make fusion power a viable future source of energy, it will be necessary to ensure that the cost of power for fusion electric generation is competitive with advanced fission concepts. In addition, fusion power will have to live up to its original promise of being a more radiologically benign technology than fission, and be able to demonstrate excellent operational safety performance. These two requirements are interrelated, since the selection of an appropriate safety philosophy early in the design phase could greatly reduce or eliminate the capital costs of elaborate safety related and protective sytems. This paper will briefly overview a few of the key safety issues presently recognized as critical to the ultimate achievement of licensable, environmentally safe and socially acceptable fusion power facilities. 12 refs

Electrical energy and cost for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Pence, G.A.

1983-01-01

An operational scenario has been developed for the Mirror Fusion Test Facility (MFTF-B) based on the System Requirements, our experience with existing systems, and discussions with the project engineers and designers who are responsible for the systems. This scenario was used to predict the amount of electrical energy needed for running the facility. A generic type listing is included for the equipment considered in each system

Ignition and Inertial Confinement Fusion at The National Ignition Facility

International Nuclear Information System (INIS)

Moses, E.

2009-01-01

The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and for studying high-energy-density (HED) science, is now operational at Lawrence Livermore National Laboratory (LLNL). The NIF is now conducting experiments to commission the laser drive, the hohlraum and the capsule and to develop the infrastructure needed to begin the first ignition experiments in FY 2010. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. NIF will achieve this by concentrating the energy from the 192 beams into a mm 3 -sized target and igniting a deuterium-tritium mix, liberating more energy than is required to initiate the fusion reaction. NIF's ignition program is a national effort managed via the National Ignition Campaign (NIC). The NIC has two major goals: execution of DT ignition experiments starting in FY2010 with the goal of demonstrating ignition and a reliable, repeatable ignition platform by the conclusion of the NIC at the end of FY2012. The NIC will also develop the infrastructure and the processes required to operate NIF as a national user facility. The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on laser fusion as a viable energy option. A laser fusion-based energy concept that builds on NIF, known as LIFE (Laser Inertial Fusion Energy), is currently under development. LIFE is inherently safe and can provide a global carbon-free energy generation solution in the 21st century. This paper describes recent progress on NIF, NIC, and the LIFE concept.

Electrical energy and cost for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Pence, G.A.

1983-02-01

An operational scenario for the Mirror Fusion Test Facility has been developed based on System Requirements, experience with existing systems, and discussions with project engineers and designers who are responsible for the systems. This scenario was used to project the electrical energy required for the facility. Each system is listed showing the equipment that has been considered, the amount of power requested, and in most cases, the power that it is now connected

Technical requirement of experiments and facilities for fusion nuclear technology

International Nuclear Information System (INIS)

Abdou, M.; Tillak, M.; Gierszwski, P.; Grover, J.; Puigh, R.; Sze, D.K.; Berwald, D.

1986-06-01

The technical issues and requirements of experiments and facilities for fusion nuclear technology (FNT) have been investigated. The nuclear subsystems addressed are: a) blanket, b) radiation shield, c) tritium processing system, and d) plasma interactive components. Emphasis has been placed on the important and complex development problems of the blanket. A technical planning process for FNT has been developed and applied, including four major elements: 1) characterization of issues, 2) quantification of testing requirements, 3) evaluation of facilities, and 4) development of a test plan to identify the role, timing, characteristics and costs of major experiments and facilities

Fusion Materials Irradiation Test Facility: experimental capabilities and test matrix

International Nuclear Information System (INIS)

Opperman, E.K.

1982-01-01

This report describes the experimental capabilities of the Fusion Materials Irradiation Test Facility (FMIT) and reference material specimen test matrices. The description of the experimental capabilities and the test matrices has been updated to match the current single test cell facility ad assessed experimenter needs. Sufficient detail has been provided so that the user can plan irradiation experiments and conceptual hardware. The types of experiments, irradiation environment and support services that will be available in FMIT are discussed

Performance test of personal RF monitor for area monitoring at magnetic confinement fusion facility

International Nuclear Information System (INIS)

Tanaka, M.; Uda, T.; Wang, J.; Fujiwara, O.

2012-01-01

For safety management at a magnetic confinement fusion-test facility, protection from not only ionising radiation, but also non-ionising radiation such as the leakage of static magnetic and electromagnetic fields is an important issue. Accordingly, the use of a commercially available personal RF monitor for multipoint area monitoring is proposed. In this study, the performance of both fast- and slow-type personal RF monitors was investigated by using a transverse electromagnetic cell system. The range of target frequencies was between 10 and 300 MHz, corresponding to the ion cyclotron range of frequency in a fusion device. The personal RF monitor was found to have good linearity, frequency dependence and isotropic response. However, the time constant for the electric field sensor of the slow-type monitor was much longer than that for the fast-type monitor. Considering the time-varying field at the facility, it is found that the fast-type monitor is suitable for multipoint monitoring at magnetic confinement fusion test facilities. (authors)

The NIF: An international high energy density science and inertial fusion user facility

Directory of Open Access Journals (Sweden)

Moses E.I.

2013-11-01

Full Text Available The National Ignition Facility (NIF, a 1.8-MJ/500-TW Nd:Glass laser facility designed to study inertial confinement fusion (ICF and high-energy-density science (HEDS, is operational at Lawrence Livermore National Laboratory (LLNL. A primary goal of NIF is to create the conditions necessary to demonstrate laboratory-scale thermonuclear ignition and burn. NIF experiments in support of indirect-drive ignition began late in FY2009 as part of the National Ignition Campaign (NIC, an international effort to achieve fusion ignition in the laboratory. To date, all of the capabilities to conduct implosion experiments are in place with the goal of demonstrating ignition and developing a predictable fusion experimental platform in 2012. The results from experiments completed are encouraging for the near-term achievement of ignition. Capsule implosion experiments at energies up to 1.6 MJ have demonstrated laser energetics, radiation temperatures, and symmetry control that scale to ignition conditions. Of particular importance is the demonstration of peak hohlraum temperatures near 300 eV with overall backscatter less than 15%. Important national security and basic science experiments have also been conducted on NIF. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of laser-driven Inertial Fusion Energy (IFE. This paper will describe the results achieved so far on the path toward ignition, the beginning of fundamental science experiments and the plans to transition NIF to an international user facility providing access to HEDS and fusion energy researchers around the world.

The NIF: An international high energy density science and inertial fusion user facility

Science.gov (United States)

Moses, E. I.; Storm, E.

2013-11-01

The National Ignition Facility (NIF), a 1.8-MJ/500-TW Nd:Glass laser facility designed to study inertial confinement fusion (ICF) and high-energy-density science (HEDS), is operational at Lawrence Livermore National Laboratory (LLNL). A primary goal of NIF is to create the conditions necessary to demonstrate laboratory-scale thermonuclear ignition and burn. NIF experiments in support of indirect-drive ignition began late in FY2009 as part of the National Ignition Campaign (NIC), an international effort to achieve fusion ignition in the laboratory. To date, all of the capabilities to conduct implosion experiments are in place with the goal of demonstrating ignition and developing a predictable fusion experimental platform in 2012. The results from experiments completed are encouraging for the near-term achievement of ignition. Capsule implosion experiments at energies up to 1.6 MJ have demonstrated laser energetics, radiation temperatures, and symmetry control that scale to ignition conditions. Of particular importance is the demonstration of peak hohlraum temperatures near 300 eV with overall backscatter less than 15%. Important national security and basic science experiments have also been conducted on NIF. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of laser-driven Inertial Fusion Energy (IFE). This paper will describe the results achieved so far on the path toward ignition, the beginning of fundamental science experiments and the plans to transition NIF to an international user facility providing access to HEDS and fusion energy researchers around the world.

Conceptual capital-cost estimate and facility design of the Mirror-Fusion Technology Demonstration Facility

International Nuclear Information System (INIS)

1982-09-01

This report contains contributions by Bechtel Group, Inc. to Lawrence Livermore National Laboratory (LLNL) for the final report on the conceptual design of the Mirror Fusion Technology Demonstration Facility (TDF). Included in this report are the following contributions: (1) conceptual capital cost estimate, (2) structural design, and (3) plot plan and plant arrangement drawings. The conceptual capital cost estimate is prepared in a format suitable for inclusion as a section in the TDF final report. The structural design and drawings are prepared as partial inputs to the TDF final report section on facilities design, which is being prepared by the FEDC

Status and possible prospects of an international fusion materials irradiation facility

International Nuclear Information System (INIS)

Cozzani, F.

1999-01-01

Structural materials for future DT fusion power reactors will have to operate under intense neutron fields with energies up to 14 MeV and fluences in the order of 2 MW/m 2 per year. As environmental acceptability, safety considerations and economic viability will be ultimately the keys to the widespread introduction of fusion power, the development of radiation-resistant and low activation materials would contribute significantly to fusion development. For this purpose, testing of materials under irradiation conditions close to those expected in a fusion power station would require the availability, in an appropriate time framework, of an intense, high-energy neutron source. Recent advances in linear accelerator technology, in small specimens testing technology, and in the comprehension of damage phenomena, lead to the conclusion that an accelerator-based D-Li neutron source, with beam energy variability, would provide the most realistic option for a fusion materials testing facility. Under the auspices of the IEA, an international effort (EU, Japan, US, RF) to carry out the conceptual design activities (CDA) of an international fusion materials irradiation facility (IFMIF), based on the D-Li concept, have been carried out successfully. A final conceptual design report was produced at the end of 1996. A phase of conceptual design evaluation (CDE), presently underway, is extending and further refining some of the conceptual design details of IFMIF. The results indicate that an IFMIF-class installation would be technically feasible and could meet its mission objectives. However, a suitable phase of Engineering Validation, to carry out some complementary R and D and prototyping, would still be needed to resolve a few key technical uncertainties before the possibility to proceed toward detailed design and construction could be explored. (orig.)

Intense neutron source facility for the fusion energy program

International Nuclear Information System (INIS)

Armstrong, D.D.; Emigh, C.R.; Meier, K.L.; Meyer, E.A.; Schneider, J.D.

1975-01-01

The Intense Neutron Source Facility, INS, has been proposed to provide a neutronic environment similar to that anticipated in a fully operational fusion-power reactor. The neutron generator will produce an intense flux of 14-MeV neutrons greater than 10 14 neutrons per cm 2 /sec from the collision of two intersecting beams, one of 1.1 A of 270 keV tritium ions and the other of a supersonic jet of deuterium gas. Using either the pure 14-MeV primary neutron spectrum or by tailoring the spectrum with appropriate moderators, crucial radiation-damage effects which are likely to occur in fusion reactors can be thoroughly explored and better understood

Fusion plasma physics research on the H-1 national facility

International Nuclear Information System (INIS)

Harris, J.

1998-01-01

Full text: Australia has a highly leveraged fusion plasma research program centred on the H-1 National Facility device at the ANU. H-1 is a heliac, a novel helical axis stellarator that was experimentally pioneered in Australia, but has a close correlation with the worldwide research program on toroidal confinement of fusion grade plasma. Experiments are conducted on H-1 by university researchers from the Australian Fusion Research Group (comprising groups from the ANU, the Universities of Sydney, Western Sydney, Canberra, New England, and Central Queensland University) under the aegis of AINSE; the scientists also collaborate with fusion researchers from Japan and the US. Recent experiments on H-1 have focused on improved confinement modes that can be accessed at very low powers in H-1, but allow the study of fundamental physics effects seen on much larger machines at higher powers. H-1 is now being upgraded in magnetic field and heating power, and will be able to confine hotter plasmas beginning in 1999, offering greatly enhanced research opportunities for Australian plasma scientists and engineers, with substantial spillover of ideas from fusion research into other areas of applied physics and engineering

FMIT - the fusion materials irradiation test facility

International Nuclear Information System (INIS)

Liska, D.J.

1980-01-01

A joint effort by the Hanford Engineering Development Laboratory (HEDL) and Los Alamos Scientific Laboratory (LASL) has produced a preliminary design for a Fusion Materials Irradiation Test Facility (FMIT) that uses a high-power linear accelerator to fire a deuteron beam into a high-speed jet of molten lithium. The result is a continuous energy spectrum of neutrons with a 14-MeV average energy which can irradiate material samples to projected end-of-life levels in about 3 years, with a total accumulated fluence of 10 21 to 10 22 n/cm 2

Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

CERN Document Server

Esposito, B; Maruccia, G; Petrizzi, L; Bignon, G; Blandin, C; Chauffriat, S; Lebrun, A; Recroix, H; Trapp, J P; Kaschuck, Y

2000-01-01

The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties...

Design of a high-flux test assembly for the Fusion Materials Irradiation Test Facility

International Nuclear Information System (INIS)

Opperman, E.K.; Vogel, M.A.

1982-01-01

The Fusion Material Test Facility (FMIT) will provide a high flux fusion-like neutron environment in which a variety of structural and non-structural materials irradiations can be conducted. The FMIT experiments, called test assemblies, that are subjected to the highest neutron flux magnitudes and associated heating rates will require forced convection liquid metal cooling systems to remove the neutron deposited power and maintain test specimens at uniform temperatures. A brief description of the FMIT facility and experimental areas is given with emphasis on the design, capabilities and handling of the high flux test assembly

Integrated assessment of thermal hydraulic processes in W7-X fusion experimental facility

Energy Technology Data Exchange (ETDEWEB)

Kaliatka, T., E-mail: tadas.kaliatka@lei.lt; Uspuras, E.; Kaliatka, A.

2017-02-15

Highlights: • The model of Ingress of Coolant Event experiment facility was developed using the RELAP5 code. • Calculation results were compared with Ingress of Coolant Event experiment data. • Using gained experience, the numerical model of Wendelstein 7-X facility was developed. • Performed analysis approved pressure increase protection system for LOCA event. - Abstract: Energy received from the nuclear fusion reaction is one of the most promising options for generating large amounts of carbon-free energy in the future. However, physical and technical problems existing in this technology are complicated. Several experimental nuclear fusion devices around the world have already been constructed, and several are under construction. However, the processes in the cooling system of the in-vessel components, vacuum vessel and pressure increase protection system of nuclear fusion devices are not widely studied. The largest amount of radioactive materials is concentrated in the vacuum vessel of the fusion device. Vacuum vessel is designed for the vacuum conditions inside the vessel. Rupture of the in-vessel components of the cooling system pipe may lead to a sharp pressure increase and possible damage of the vacuum vessel. To prevent the overpressure, the pressure increase protection system should be designed and implemented. Therefore, systematic and detailed experimental and numerical studies, regarding the thermal-hydraulic processes in cooling system, vacuum vessel and pressure increase protection system, are important and relevant. In this article, the numerical investigation of thermal-hydraulic processes in cooling systems of in-vessel components, vacuum vessels and pressure increase protection system of fusion devices is presented. Using the experience gained from the modelling of “Ingress of Coolant Event” experimental facilities, the numerical model of Wendelstein 7-X (W7-X) experimental fusion device was developed. The integrated analysis of the

All Solid State Optical Pulse Shaper for the OMEGA Laser Fusion Facility

International Nuclear Information System (INIS)

Okishev, A.V.; Skeldon, M.D.; Keck, R.L.; Seka, W.

2000-01-01

OAK-B135 All Solid State Optical Pulse Shaper for the OMEGA Laser Fusion Facility. The authors have developed an all-solid-state, compact, computer-controlled, flexible optical pulse shaper for the OMEGA laser facility. This pulse shaper produces high bandwidth, temporally shaped laser pulses that meet OMEGA requirements. The design is a significant simplification over existing technology with improved performance capabilities

Magnet Design Considerations for Fusion Nuclear Science Facility

Energy Technology Data Exchange (ETDEWEB)

Zhai, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Kessel, C. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison, WI (United States) Fusion Technology Institute; Titus, P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

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 7.5 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.

Advances in inertial confinement fusion at the National Ignition Facility (NIF)

International Nuclear Information System (INIS)

Moses, Edward I.

2010-01-01

The 192-beam National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in Livermore, CA, is now operational and conducting experiments. NIF, the flagship facility of the U.S. Inertial Confinement Fusion (ICF) Program, will achieve high-energy-density conditions never previously obtained in the laboratory-temperatures over 100 million K, densities of 1000 g/cm 3 , and pressures exceeding 100 billion atmospheres. Such conditions exist naturally only in the interiors of the stars and during thermonuclear burn. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. To date, the NIF laser has demonstrated all pulse shape, beam quality, energy, and other specifications required to meet the ignition challenge. On March 10, 2009, the NIF laser delivered 1.1 MJ of ultraviolet laser energy to target chamber center, approximately 30 times more energy than any previous facility. The ignition program at NIF is the National Ignition Campaign (NIC), a national collaboration for ignition experimentation with participation from General Atomics, LLNL, Los Alamos National Laboratory (LANL), Sandia National Laboratories (SNL), and the University of Rochester Laboratory for Laser Energetics (LLE). The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on fusion as a viable energy option. A particular energy concept under investigation is the LIFE (Laser Inertial Fusion Energy) scheme. The LIFE engine is inherently safe, minimizes proliferation concerns associated with the nuclear fuel cycle, and can provide a sustainable carbon-free energy generation solution in the 21st century. This talk will describe NIF and its potential as a user facility and an experimental platform for high-energy-density science, NIC, and the LIFE approach for clean, sustainable energy.

Advances in Inertial Confinement Fusion at the National Ignition Facility (NIF)

International Nuclear Information System (INIS)

Moses, E.

2009-01-01

The 192-beam National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in Livermore, CA, is now operational and conducting experiments. NIF, the flagship facility of the U.S. Inertial Confinement Fusion (ICF) Program, will achieve high-energy-density conditions never previously obtained in the laboratory - temperatures over 100 million K, densities of 1,000 g/cm3, and pressures exceeding 100 billion atmospheres. Such conditions exist naturally only in the interiors of the stars and during thermonuclear burn. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. To date, the NIF laser has demonstrated all pulse shape, beam quality, energy, and other specifications required to meet the ignition challenge. On March 10, 2009, the NIF laser delivered 1.1 MJ of ultraviolet laser energy to target chamber center, approximately 30 times more energy than any previous facility. The ignition program at NIF is the National Ignition Campaign (NIC), a national collaboration for ignition experimentation with participation from General Atomics, LLNL, Los Alamos National Laboratory (LANL), Sandia National Laboratories (SNL), and the University of Rochester Laboratory for Laser Energetics (LLE). The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on fusion as a viable energy option. A particular energy concept under investigation is the LIFE (Laser Inertial Fusion Energy) scheme. The LIFE engine is inherently safe, minimizes proliferation concerns associated with the nuclear fuel cycle, and can provide a sustainable carbon-free energy generation solution in the 21st century. This talk will describe NIF and its potential as a user facility and an experimental platform for high-energy-density science, NIC, and the LIFE approach for clean, sustainable energy.

Operational characteristics of the OMEGA short-wavelength laser fusion facility

International Nuclear Information System (INIS)

Soures, J.M.; Hutchison, R.; Jacobs, S.; McCrory, R.L.; Peck, R.; Seka, W.

1984-01-01

Twelve beams of the OMEGA, 24 beam direct-drive laser facility have been converted to 351-nm wavelength operation. The performance characteristics of this short-wavelength facility will be discussed. Beam-to-beam energy balance of +-2.3% and on-target energy, at 351-nm, in excess of 70 J per beam have been demonstrated. Long-term performance (>600 shots) of the system has been optimized by appropriate choice of index matching liquid, optical materials and coatings. The application of this system in direct-drive laser fusion experiments will be discussed

Physical and engineering aspects of a fusion engineering test facility based on mirror confinement

International Nuclear Information System (INIS)

Kawabe, T.; Hirayama, S.; Hojo, H.; Kozaki, Y.; Yoshikawa, K.

1986-01-01

Controlled fusion research has accomplished great progress in the field of confinement of high-density and high-temperature plasmas and breakeven experiments are expected before the end of the 1980s. Many experiments have been proposed as the next step for fusion research. Among them is the study of ignited plasmas and another is the study of fusion engineering. Some of the important studies in fusion engineering are the integrated test in a fusion reactor environment as well as tests of first-wall materials and of the reactor structures, and test for tritium breeding and blanket modules or submodules. An ideal neutron source for the study of fusion engineering is the deuterium-tritium (D-T) fusion plasma itself. A neutron facility based on a D-T-burning plasma consists of all of the components that a real fusion power reactor would have, so eventually the integrated test for fusion reactor engineering can be done as well as the tests for each engineering component

Optimization of the testing volumes with respect to neutron flux levels in the two-target high flux D-Li neutron source for the international fusion materials irradiation facility

International Nuclear Information System (INIS)

Kelleher, W.P.; Varsamis, G.L.

1989-01-01

An economic and fusion-relevant source of high-energy neutrons is an essential element in the fusion nuclear technology and development program. This source can be generated by directing a high energy deuteron beam onto a flowing liquid lithium target, producing neutrons via the D-Lithium stripping reaction. Previous work on this type of source concentrated on a design employing one deuteron beam of modest amperage. This design was shown to have a relatively small testing volume with high flux gradients and was therefor considered somewhat unattractive from a materials testing standpoint. A design using two lithium targets and two high-amperage beams has recently been proposed. This two beam design has been examined in an effort to maximize the test volume while minimizing the flux gradients and minimizing the effect of radiation damage on one target due to the other. A spatial, energy and angle dependent neutron source modeling the D-Lithium source was developed. Using this source, a 3-dimensional map of uncollided flux within the test volume was calculated. The results showed that the target separation has little effect on the available experimental volume and that a testing volume of ∼35 liters is available with a volume averaged flux above 10 14 n/cm 2 /s. The collided flux within the test volume was then determined by coupling the source model with a Monte Carlo code. The spectral effects of the high-energy tail in the flux were examined and evaluated as to possible effects on materials response. Calculations comparing the radiation damage to materials from the D-Lithium source to that cause by a standard DT fusion first-wall neutron flux spectrum showed that the number of appm and dpa, as well as the ratio appm/dpa and dpa/MW/m 2 are within 30% for the two sources. 8 refs., 8 figs

Vacuum vessel for the tandem Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Gerich, J.W.

1986-01-01

In 1980, the US Department of Energy gave the Lawrence Livermore National Laboratory approval to design and build a tandem Mirror Fusion Test Facility (MFTF-B) to support the goals of the National Mirror Program. We designed the MFTF-B vacuum vessel both to maintain the required ultrahigh vacuum environment and to structurally support the 42 superconducting magnets plus auxiliary internal and external equipment. During our design work, we made extensive use of both simple and complex computer models to arrive at a cost-effective final configuration. As part of this work, we conducted a unique dynamic analysis to study the interaction of the 32,000-tonne concrete-shielding vault with the 2850-tonne vacuum vessel system. To maintain a vacuum of 2 x 10 -8 torr during the physics experiments inside the vessel, we designed a vacuum pumping system of enormous capacity. The vacuum vessel (4200-m 3 internal volume) has been fabricated and erected, and acceptance tests have been completed at the Livermore site. The rest of the machine has been assembled, and individual systems have been successfully checked. On October 1, 1985, we began a series of integrated engineering tests to verify the operation of all components as a complete system

Nuclear data for the production of radioisotopes in fusion materials irradiation facility

International Nuclear Information System (INIS)

Cheng, E.T.; Schenter, R.E.; Mann, F.M.; Ikeda, Y.

1991-01-01

The fusion materials irradiation facility (FMIF) is a neutron source generator that will produce a high-intensity 14-MeV neutron field for testing candidate fusion materials under reactor irradiation conditions. The construction of such a facility is one of the very important development stages toward realization of fusion energy as a practical energy source for electricity production. As a result of the high-intensity neutron field, 10 MW/m 2 or more equivalent neutron wall loading, and the relatively high-energy (10- to 20-MeV) neutrons, the FMIF, as future fusion reactors, also bears the potential capability of producing a significant quantity of radioisotopes. A study is being conducted to identify the potential capability of the FMIF to produce radioisotopes for medical and industrial applications. Two types of radioisotopes are involved: one is already available; the second might not be readily available using conventional production methods. For those radioisotopes that are not readily available, the FMIF could develop significant benefits for future generations as a result of the availability of such radioisotopes for medical or industrial applications. The current production of radioisotopes could help finance the operation of the FMIF for irradiating the candidate fusion materials; thus this concept is attractive. In any case, nuclear data are needed for calculating the neutron flux and spectrum in the FMIF and the potential production rates of these isotopes. In this paper, the authors report the result of a preliminary investigation on the production of 99 Mo, the parent radioisotope for 99m Tc

Fusion facility siting considerations

International Nuclear Information System (INIS)

Bussell, G.T.

1985-01-01

Inherent in the fusion program's transition from hydrogen devices to commercial power machines is a general increase in the size and scope of succeeding projects. This growth will lead to increased emphasis on safety, environmental impact, and the external effects of fusion in general, and of each new device in particular. A critically important consideration in this regard is site selection. The purpose of this paper is to examine major siting issues that may affect the economics, safety, and environmental impact of fusion

Overview of recent AWE fusion-related studies, experiments and facilities

Directory of Open Access Journals (Sweden)

Roberts P.D.

2013-11-01

Full Text Available The presentation will describe the current status of modelling short and long pulse laser irradiation and its application to inertial fusion designs. Recent results will be described which give confidence in the modelling in specific regimes. An update will be given of the AWE ORION laser facility and the availability planned for academic access.

Large power supply facilities for fusion research

International Nuclear Information System (INIS)

Miyahara, Akira; Yamamoto, Mitsuyoshi.

1976-01-01

The authors had opportunities to manufacture and to operate two power supply facilities, that is, 125MVA computer controlled AC generator with a fly wheel for JIPP-T-2 stellerator in Institute of Plasma Physics, Nagoya University and 3MW trial superconductive homopolar DC generator to the Japan Society for Promotion of Machine Industry. The 125MVA fly-wheel generator can feed both 60MW (6kV x 10kA) DC power for toroidal coils and 20MW (0.5kV x 40kA) DC power for helical coils. The characteristic features are possibility of Bung-Bung control based on Pontrjagin's maximum principle, constant current control or constant voltage control for load coils, and cpu control for routine operation. The 3MW (150V-20000A) homopolar generator is the largest in the world as superconductive one, however, this capacity is not enough for nuclear fusion research. The problems of power supply facilities for large Tokamak devices are discussed

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

Science.gov (United States)

Chan, V. S.; Costley, A. E.; Wan, B. N.; Garofalo, A. M.; Leuer, J. A.

2015-02-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher BT and Ip can result in a high gain Qfus ˜ 12, Pfus ˜ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target.

Nuclear engineering questions: power, reprocessing, waste, decontamination, fusion

International Nuclear Information System (INIS)

Walton, R.D. Jr.

1979-01-01

This volume contains papers presented at the chemical engineering symposium on nuclear questions. Specific questions addressed by the speakers included: nuclear power - why and how; commercial reprocessing - permanent death or resurrection; long-term management of commercial high-level wastes; long-term management of defense high-level waste; decontamination and decommissioning of nuclear facilities, engineering aspects of laser fusion I; and engineering aspects of laser fusion II. Individual papers have been input to the Energy Data Base previously

International Fusion Materials Irradiation Facility conceptual design activity. Present status and perspective

International Nuclear Information System (INIS)

Kondo, Tatsuo; Noda, Kenji; Oyama, Yukio

1998-01-01

For developing the materials for nuclear fusion reactors, it is indispensable to study on the neutron irradiation behavior under fusion reactor conditions, but there is not any high energy neutron irradiation facility that can simulate fusion reactor conditions at present. Therefore, the investigation of the IFMIF was begun jointly by Japan, USA, Europe and Russia following the initiative of IEA. The conceptual design activities were completed in 1997. As to the background and the course, the present status of the research on heavy irradiation and the testing means for fusion materials, the requirement and the technical basis of high energy neutron irradiation, and the international joint design activities are reported. The materials for fusion reactors are exposed to the neutron irradiation with the energy spectra up to 14 MeV. The requirements from the users that the IFMIF should satisfy, the demand of the tests for the materials of prototype and demonstration fusion reactors and the evaluation of the neutron field characteristics of the IFMIF are discussed. As to the conceptual design of the IFMIF, the whole constitution, the operational mode, accelerator system and target system are described. (K.I.)

Safety issues to be taken into account in designing future nuclear fusion facilities

Energy Technology Data Exchange (ETDEWEB)

Perrault, Didier, E-mail: didier.perrault@irsn.fr

2016-11-01

Highlights: • Assess if decay heat removal is a safety function. • Re-study accidents considered for ITER and identify those specific to DEMO. • Limit tritium inventory and optimize main gaseous tritium release routes. • Take into account constraints related to requirements of waste disposal routes. - Abstract: For several years now, the French “Institut de Radioprotection et de Sûreté Nucléaire” has been carrying out expertise of ITER fusion facility safety files at the request of the French “Autorité de Sûreté Nucléaire”. As part of the lengthy process which should lead to mastering nuclear fusion, different fusion facility projects are currently under study throughout the world to be ready to continue building on the work which will take place in the ITER facility. On the basis of the experience acquired during the ITER safety expertise, the IRSN has carried out a preliminary study of the safety issues which seem necessary to take into account right from the earliest design phase of these DEMO facilities. The issues studied have included the decay heat removal, exposure to ionizing radiation, potential accidents, and effluent releases and waste. The study shows that it will be important to give priority to the following actions, given that their results would have a major influence on the design: assess if decay heat removal is a safety function, re-study the accidents considered in the context of the ITER project and identify those specific to DEMO, and optimize each of the main routes for gaseous tritium releases.

Diagnostic technique for measuring fusion reaction rate for inertial confinement fusion experiments at Shen Guang-III prototype laser facility

International Nuclear Information System (INIS)

Wang Feng; Peng Xiao-Shi; Liu Shen-Ye; Xu Tao; Kang Dong-Guo

2013-01-01

A study is conducted using a two-dimensional simulation program (Lared-s) with the goal of developing a technique to evaluate the effect of Rayleigh-Taylor growth in a neutron fusion reaction region. Two peaks of fusion reaction rate are simulated by using a two-dimensional simulation program (Lared-s) and confirmed by the experimental results. A neutron temporal diagnostic (NTD) system is developed with a high temporal resolution of ∼ 30 ps at the Shen Guang-III (SG-III) prototype laser facility in China, to measure the fusion reaction rate history. With the shape of neutron reaction rate curve and the spherical harmonic function in this paper, the degree of Rayleigh-Taylor growth and the main source of the neutron yield in our experiment can be estimated qualitatively. This technique, including the diagnostic system and the simulation program, may provide important information for obtaining a higher neutron yield in implosion experiments of inertial confinement fusion

Directions for possible upgrades of the Mirror Fusion Test Facility (MFTF)

International Nuclear Information System (INIS)

Damm, C.C.; Coensgen, F.H.; Devoto, R.S.; Molvik, A.W.; Porter, G.D.; Shearer, J.W.; Stallard, B.W.

1977-01-01

The Mirror Fusion Test Facility (MFTF) may be upgraded by extending the time of plasma sustenance in an approach to steady-state operation and/or by increasing the neutral-beam injection energy. Some parameter bounds for these upgrades are discussed as they relate to a definition of the required neutral-beam development

Recent developments in IFE safety and tritium research and considerations for future nuclear fusion facilities

International Nuclear Information System (INIS)

Reyes, Susana; Anklam, Tom; Meier, Wayne; Campbell, Patrick; Babineau, Dave; Becnel, James; Taylor, Craig; Coons, Jim

2016-01-01

Highlights: • The safety characteristics and at risk inventories in an IFE facility are discussed. • The primary nuclear hazard is the potential exposure of workers and/or the public to tritium and/or neutronically activated products. • Recent technology developments in tritium processing are key for minimization of inventories. • Initial safety studies indicate that hazards associated to the use of liquid lithium can be appropriately managed. • Simulation of worst-case scenarios indicate that the accident consequences are limited and below the limit for public evacuation. - Abstract: Over the past five years, the fusion energy group at Lawrence Livermore National Laboratory (LLNL) has made significant progress in the area of safety and tritium research for Inertial Fusion Energy (IFE). Focus has been driven towards the minimization of inventories, accident safety, development of safety guidelines and licensing considerations. Recent technology developments in tritium processing and target fill have had a major impact on reduction of tritium inventories in the facility. A safety advantage of inertial fusion energy using indirect-drive targets is that the structural materials surrounding the fusion reactions can be protected from target emissions by a low-pressure chamber fill gas, therefore eliminating plasma-material erosion as a source of activated dust production. An important inherent safety advantage of IFE when compared to other magnetic fusion energy (MFE) concepts that have been proposed to-date (including ITER), is that loss of plasma control events with the potential to damage the first wall, such as disruptions, are non-conceivable, therefore eliminating a number of potential accident initiators and radioactive in-vessel source term generation. In this paper, we present an overview of the safety assessments performed to-date, comparing results to the US DOE Fusion Safety Standards guidelines and the recent lessons-learnt from ITER safety and

Recent developments in IFE safety and tritium research and considerations for future nuclear fusion facilities

Energy Technology Data Exchange (ETDEWEB)

Reyes, Susana, E-mail: reyes20@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Anklam, Tom; Meier, Wayne; Campbell, Patrick [Lawrence Livermore National Laboratory, Livermore, CA (United States); Babineau, Dave; Becnel, James [Savannah River National Laboratory, Aiken, SC (United States); Taylor, Craig; Coons, Jim [Los Alamos National Laboratory, Los Alamos, NM (United States)

2016-11-01

Highlights: • The safety characteristics and at risk inventories in an IFE facility are discussed. • The primary nuclear hazard is the potential exposure of workers and/or the public to tritium and/or neutronically activated products. • Recent technology developments in tritium processing are key for minimization of inventories. • Initial safety studies indicate that hazards associated to the use of liquid lithium can be appropriately managed. • Simulation of worst-case scenarios indicate that the accident consequences are limited and below the limit for public evacuation. - Abstract: Over the past five years, the fusion energy group at Lawrence Livermore National Laboratory (LLNL) has made significant progress in the area of safety and tritium research for Inertial Fusion Energy (IFE). Focus has been driven towards the minimization of inventories, accident safety, development of safety guidelines and licensing considerations. Recent technology developments in tritium processing and target fill have had a major impact on reduction of tritium inventories in the facility. A safety advantage of inertial fusion energy using indirect-drive targets is that the structural materials surrounding the fusion reactions can be protected from target emissions by a low-pressure chamber fill gas, therefore eliminating plasma-material erosion as a source of activated dust production. An important inherent safety advantage of IFE when compared to other magnetic fusion energy (MFE) concepts that have been proposed to-date (including ITER), is that loss of plasma control events with the potential to damage the first wall, such as disruptions, are non-conceivable, therefore eliminating a number of potential accident initiators and radioactive in-vessel source term generation. In this paper, we present an overview of the safety assessments performed to-date, comparing results to the US DOE Fusion Safety Standards guidelines and the recent lessons-learnt from ITER safety and

Design of a fusion engineering test facility

International Nuclear Information System (INIS)

Sager, P.H.

1980-01-01

The fusion Engineering Test Facility (ETF) is being designed to provide for engineering testing capability in a program leading to the demonstration of fusion as a viable energy option. It will combine power-reactor-type components and subsystems into an integrated tokamak system and provide a test bed to test blanket modules in a fusion environment. Because of the uncertainties in impurity control two basic designs are being developed: a design with a bundle divertor (Design 1) and one with a poloidal divertor (Design 2). The two designs are similar where possible, the latter having somewhat larger toroidal field (TF) coils to accommodate removal of the larger torus sectors required for the single-null poloidal divertor. Both designs have a major radius of 5.4 m, a minor radius of 1.3 m, and a D-shaped plasma with an elongation of 1.6. Ten TF coils are incorporated in both designs, producing a toroidal field of 5.5 T on-axis. The ohmic heating and equilibrium field (EF) coils supply sufficient volt-seconds to produce a flat-top burn of 100 s and a duty cycle of 135 s, including a start of 12 s, a burn termination of 10 s, and a pumpdown of 13 s. The total fusion power during burn is 750 MW, giving a neutron wall loading of 1.5 MW/m 2 . In Design 1 of the poloidal field (PF) coils except the fast-response EF coils are located outside the FT coils and are superconducting. The fast-response coils are located inside the TF coil bore near the torus and are normal conducting so that they can be easily replaced.In Design 2 all of the PF coils are located outside the TF coils and are superconducting. Ignition is achieved with 60 MW of neutral beam injection at 150 keV. Five megawatts of radio frequency heating (electron cyclotron resonance heating) is used to assist in the startup and limit the breakdown requirement to 25 V

Study on system integration of robots operated in nuclear fusion facility and nuclear power plant facilities

International Nuclear Information System (INIS)

Oka, Kiyoshi

2004-07-01

A present robot is required to apply to many fields such as amusement, welfare and protection against disasters. The are however only limited numbers of the robots, which can work under the actual conditions as a robot system. It is caused by the following reasons: (1) the robot system cannot be realized by the only collection of the elemental technologies, (2) the performance of the robot is determined by that of the integrated system composed of the complicated elements with many functions, and (3) the respective elements have to be optimized in the integrated robot system with a well balance among them, through their examination, adjustment and improvement. Therefore, the system integration of the robot composed of a large number of elements is the most critical issue to realize the robot system for actual use. In the present paper, I describe the necessary approaches and elemental technologies to solve the issues on the system integration of the typical robot systems for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities. These robots work under the intense radiation condition and restricted space in place of human. In particular, I propose a new approach to realize the system integration of the robot for actual use from the viewpoints of not only the environment and working conditions but also the restructure and optimization of the required elemental technologies with a well balance in the robot system. Based on the above approach, I have a contribution to realize the robot systems working under the actual conditions for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities. (author)

Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes

International Nuclear Information System (INIS)

Chan, V.S.; Garofalo, A.M.; Leuer, J.A.; Costley, A.E.; Wan, B.N.

2015-01-01

This paper presents the results of a multi-system codes benchmarking study of the recently published China Fusion Engineering Test Reactor (CFETR) pre-conceptual design (Wan et al 2014 IEEE Trans. Plasma Sci. 42 495). Two system codes, General Atomics System Code (GASC) and Tokamak Energy System Code (TESC), using different methodologies to arrive at CFETR performance parameters under the same CFETR constraints show that the correlation between the physics performance and the fusion performance is consistent, and the computed parameters are in good agreement. Optimization of the first wall surface for tritium breeding and the minimization of the machine size are highly compatible. Variations of the plasma currents and profiles lead to changes in the required normalized physics performance, however, they do not significantly affect the optimized size of the machine. GASC and TESC have also been used to explore a lower aspect ratio, larger volume plasma taking advantage of the engineering flexibility in the CFETR design. Assuming the ITER steady-state scenario physics, the larger plasma together with a moderately higher B T and I p can result in a high gain Q fus  ∼ 12, P fus  ∼ 1 GW machine approaching DEMO-like performance. It is concluded that the CFETR baseline mode can meet the minimum goal of the Fusion Nuclear Science Facility (FNSF) mission and advanced physics will enable it to address comprehensively the outstanding critical technology gaps on the path to a demonstration reactor (DEMO). Before proceeding with CFETR construction steady-state operation has to be demonstrated, further development is needed to solve the divertor heat load issue, and blankets have to be designed with tritium breeding ratio (TBR) >1 as a target. (paper)

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA; Volume 5

Energy Technology Data Exchange (ETDEWEB)

Dionne, B.J.; Sullivan, S.G.; Baum, J.W. [Brookhaven National Lab., Upton, NY (United States)

1994-01-01

Promoting the exchange of information related to implementation of the As Low as Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report was prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health. It contains the fifth in a series of bibliographies on dose reduction at DOE facilities. The BNL ALARA Center was originally established in 1983 under the sponsorship of the Nuclear Regulatory Commission to monitor dose-reduction research and ALARA activities at nuclear power plants. This effort was expanded in 1988 by the DOE`s Office of Environment, Safety and Health, to include DOE nuclear facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and accelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts.

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA; Volume 5

International Nuclear Information System (INIS)

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1994-01-01

Promoting the exchange of information related to implementation of the As Low as Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report was prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health. It contains the fifth in a series of bibliographies on dose reduction at DOE facilities. The BNL ALARA Center was originally established in 1983 under the sponsorship of the Nuclear Regulatory Commission to monitor dose-reduction research and ALARA activities at nuclear power plants. This effort was expanded in 1988 by the DOE's Office of Environment, Safety and Health, to include DOE nuclear facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and accelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts

Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Perry, E.; Chrzanowski, J.; Rule, K.; Viola, M.; Williams, M.; Strykowsky, R.

1999-01-01

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D and D) of the TFTR is scheduled to occur over a period of three years beginning in October 1999. This is not a typical Department of Energy D and D Project where a facility is isolated and cleaned up by ''bulldozing'' all facility and hardware systems to a greenfield condition. The mission of TFTR D and D is to: (a) surgically remove items which can be re-used within the DOE complex, (b) remove tritium contaminated and activated systems for disposal, (c) clear the test cell of hardware for future reuse, (d) reclassify the D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Facility Safety) and (e) provide data on the D and D of a large magnetic fusion facility. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The record-breaking deuterium-tritium experiments performed on TFTR resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 75 mRem/hr. These radiological hazards along with the size and shape of the Tokamak present a unique and challenging task for dismantling

Reduced cost design of liquid lithium target for international fusion material irradiation facility (IFMIF)

International Nuclear Information System (INIS)

Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki

2001-01-01

The International Fusion Materials Irradiation Facility (IFMIF) is being jointly planned to provide an accelerator-based D-Li neutron source to produce intense high energy neutrons (2 MW/m 2 ) up to 200 dpa and a sufficient irradiation volume (500 cm 3 ) for testing the candidate materials and components up to about a full lifetime of their anticipated use in ITER and DEMO. To realize such a condition, 40 MeV deuteron beam with a current of 250 mA is injected into high speed liquid lithium flow with a speed of 20 m/s. Following Conceptual Design Activity (1995-1998), a design study with focus on cost reduction without changing its original mission has been done in 1999. The following major changes to the CAD target design have been considered in the study and included in the new design: i) number of the Li target has been changed from 2 to 1, ii) spare of impurity traps of the Li loop was removed although the spare will be stored in a laboratory for quick exchange, iii) building volume was reduced via design changes in lithium loop length. This paper describes the reduced cost design of the lithium target system and recent status of Key Element Technology activities. (author)

Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

International Nuclear Information System (INIS)

Beller, D.E.

1986-01-01

A new conceptual design of a fusion reactor blanket simulation facility was developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBR), because experiments conducted in it have resulted in the discovery of deficiencies in neutronics prediction methods. With this design, discrepancies between calculation and experimental data can be fully attributed to calculation methods because design deficiencies that could affect results are insignificant. Inelastic scattering cross sections are identified as a major source of these discrepancies. The conceptual design of this FBBR analog, the fusion reactor blanket facility (FRBF), is presented. Essential features are a cylindrical geometry and a distributed, cosine-shaped line source of 14-MeV neutrons. This source can be created by sweeping a deuteron beam over an elongated titanium-tritide target. To demonstrate that the design of the FRBF will not contribute significant deviations in experimental results, neutronics analyses were performed: results of comparisons of 2-dimensional to 1-dimensional predictions are reported for two blanket compositions. Expected deviations from 1-D predictions which are due to source anisotropy and blanket asymmetry are minimal. Thus, design of the FRBF allows simple and straightforward interpretation of the experimental results, without a need for coarse 3-D calculations

Accuracy of volume measurement using 3D ultrasound and development of CT-3D US image fusion algorithm for prostate cancer radiotherapy

International Nuclear Information System (INIS)

Baek, Jihye; Huh, Jangyoung; Hyun An, So; Oh, Yoonjin; Kim, Myungsoo; Kim, DongYoung; Chung, Kwangzoo; Cho, Sungho; Lee, Rena

2013-01-01

Purpose: To evaluate the accuracy of measuring volumes using three-dimensional ultrasound (3D US), and to verify the feasibility of the replacement of CT-MR fusion images with CT-3D US in radiotherapy treatment planning. Methods: Phantoms, consisting of water, contrast agent, and agarose, were manufactured. The volume was measured using 3D US, CT, and MR devices. A CT-3D US and MR-3D US image fusion software was developed using the Insight Toolkit library in order to acquire three-dimensional fusion images. The quality of the image fusion was evaluated using metric value and fusion images. Results: Volume measurement, using 3D US, shows a 2.8 ± 1.5% error, 4.4 ± 3.0% error for CT, and 3.1 ± 2.0% error for MR. The results imply that volume measurement using the 3D US devices has a similar accuracy level to that of CT and MR. Three-dimensional image fusion of CT-3D US and MR-3D US was successfully performed using phantom images. Moreover, MR-3D US image fusion was performed using human bladder images. Conclusions: 3D US could be used in the volume measurement of human bladders and prostates. CT-3D US image fusion could be used in monitoring the target position in each fraction of external beam radiation therapy. Moreover, the feasibility of replacing the CT-MR image fusion to the CT-3D US in radiotherapy treatment planning was verified.

Designing of High-Volume PET/CT Facility with Optimal Reduction of Radiation Exposure to the Staff: Implementation and Optimization in a Tertiary Health Care Facility in India

International Nuclear Information System (INIS)

Jha, Ashish Kumar; Singh, Abhijith Mohan; Mithun, Sneha; Shah, Sneha; Agrawal, Archi; Purandare, Nilendu C.; Shetye, Bhakti; Rangarajan, Venkatesh

2015-01-01

Positron emission tomography (PET) has been in use for a few decades but with its fusion with computed tomography (CT) in 2001, the new PET/CT integrated system has become very popular and is now a key influential modality for patient management in oncology. However, along with its growing popularity, a growing concern of radiation safety among the radiation professionals has become evident. We have judiciously developed a PET/CT facility with optimal shielding, along with an efficient workflow to perform high volume procedures and minimize the radiation exposure to the staff and the general public by reducing unnecessary patient proximity to the staff and general public

Safety Culture and Best Practices at Japan's Fusion Research Facilities

Energy Technology Data Exchange (ETDEWEB)

Rule, Keith [PPPL

2014-05-01

The Safety Monitor Joint Working Group (JWG) is one of the magnetic fusion research collaborations between the US Department of Energy and the government of Japan. Visits by occupational safety personnel are made to participating institutions on a biennial basis. In the 2013 JWG visit of US representatives to Japan, the JWG members noted a number of good safety practices in the safety walkthroughs. These good practices and safety culture topics are discussed in this paper. The JWG hopes that these practices for worker safety can be adopted at other facilities. It is a well-known, but unquantified, safety principle that well run, safe facilities are more productive and efficient than other facilities (Rule, 2009). Worker safety, worker productivity, and high quality in facility operation all complement each other (Mottel, 1995).

The Terra Data Fusion Project: An Update

Science.gov (United States)

Di Girolamo, L.; Bansal, S.; Butler, M.; Fu, D.; Gao, Y.; Lee, H. J.; Liu, Y.; Lo, Y. L.; Raila, D.; Turner, K.; Towns, J.; Wang, S. W.; Yang, K.; Zhao, G.

2017-12-01

Terra is the flagship of NASA's Earth Observing System. Launched in 1999, Terra's five instruments continue to gather data that enable scientists to address fundamental Earth science questions. By design, the strength of the Terra mission has always been rooted in its five instruments and the ability to fuse the instrument data together for obtaining greater quality of information for Earth Science compared to individual instruments alone. As the data volume grows and the central Earth Science questions move towards problems requiring decadal-scale data records, the need for data fusion and the ability for scientists to perform large-scale analytics with long records have never been greater. The challenge is particularly acute for Terra, given its growing volume of data (> 1 petabyte), the storage of different instrument data at different archive centers, the different file formats and projection systems employed for different instrument data, and the inadequate cyberinfrastructure for scientists to access and process whole-mission fusion data (including Level 1 data). Sharing newly derived Terra products with the rest of the world also poses challenges. As such, the Terra Data Fusion Project aims to resolve two long-standing problems: 1) How do we efficiently generate and deliver Terra data fusion products? 2) How do we facilitate the use of Terra data fusion products by the community in generating new products and knowledge through national computing facilities, and disseminate these new products and knowledge through national data sharing services? Here, we will provide an update on significant progress made in addressing these problems by working with NASA and leveraging national facilities managed by the National Center for Supercomputing Applications (NCSA). The problems that we faced in deriving and delivering Terra L1B2 basic, reprojected and cloud-element fusion products, such as data transfer, data fusion, processing on different computer architectures

Ignition on the National Ignition Facility: a path towards inertial fusion energy

International Nuclear Information System (INIS)

Moses, Edward I.

2009-01-01

The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and experiments studying high-energy-density (HED) science, is nearing completion at Lawrence Livermore National Laboratory (LLNL). NIF, a 192-beam Nd-glass laser facility, will produce 1.8 MJ, 500 TW of light at the third-harmonic, ultraviolet light of 351 nm. The NIF project is scheduled for completion in March 2009. Currently, all 192 beams have been operationally qualified and have produced over 4.0 MJ of light at the fundamental wavelength of 1053 nm, making NIF the world's first megajoule laser. The principal goal of NIF is to achieve ignition of a deuterium-tritium (DT) fuel capsule and provide access to HED physics regimes needed for experiments related to national security, fusion energy and for broader scientific applications. The plan is to begin 96-beam symmetric indirect-drive ICF experiments early in FY2009. These first experiments represent the next phase of the National Ignition Campaign (NIC). This national effort to achieve fusion ignition is coordinated through a detailed plan that includes the science, technology and equipment such as diagnostics, cryogenic target manipulator and user optics required for ignition experiments. Participants in this effort include LLNL, General Atomics, Los Alamos National Laboratory, Sandia National Laboratory and the University of Rochester Laboratory for Energetics (LLE). The primary goal for NIC is to have all of the equipment operational and integrated into the facility soon after project completion and to conduct a credible ignition campaign in 2010. When the NIF is complete, the long-sought goal of achieving self-sustaining nuclear fusion and energy gain in the laboratory will be much closer to realization. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of inertial fusion energy (IFE) and will likely focus

Ignition on the National Ignition Facility: a path towards inertial fusion energy

Science.gov (United States)

Moses, Edward I.

2009-10-01

The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and experiments studying high-energy-density (HED) science, is nearing completion at Lawrence Livermore National Laboratory (LLNL). NIF, a 192-beam Nd-glass laser facility, will produce 1.8 MJ, 500 TW of light at the third-harmonic, ultraviolet light of 351 nm. The NIF project is scheduled for completion in March 2009. Currently, all 192 beams have been operationally qualified and have produced over 4.0 MJ of light at the fundamental wavelength of 1053 nm, making NIF the world's first megajoule laser. The principal goal of NIF is to achieve ignition of a deuterium-tritium (DT) fuel capsule and provide access to HED physics regimes needed for experiments related to national security, fusion energy and for broader scientific applications. The plan is to begin 96-beam symmetric indirect-drive ICF experiments early in FY2009. These first experiments represent the next phase of the National Ignition Campaign (NIC). This national effort to achieve fusion ignition is coordinated through a detailed plan that includes the science, technology and equipment such as diagnostics, cryogenic target manipulator and user optics required for ignition experiments. Participants in this effort include LLNL, General Atomics, Los Alamos National Laboratory, Sandia National Laboratory and the University of Rochester Laboratory for Energetics (LLE). The primary goal for NIC is to have all of the equipment operational and integrated into the facility soon after project completion and to conduct a credible ignition campaign in 2010. When the NIF is complete, the long-sought goal of achieving self-sustaining nuclear fusion and energy gain in the laboratory will be much closer to realization. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of inertial fusion energy (IFE) and will likely focus

The NASA-Lewis program on fusion energy for space power and propulsion, 1958-1978

International Nuclear Information System (INIS)

Schulze, N.R.; Roth, J.R.

1990-01-01

An historical synopsis is provided of the NASA-Lewis research program on fusion energy for space power and propulsion systems. It was initiated to explore the potential applications of fusion energy to space power and propulsion systems. Some fusion related accomplishments and program areas covered include: basic research on the Electric Field Bumpy Torus (EFBT) magnetoelectric fusion containment concept, including identification of its radial transport mechanism and confinement time scaling; operation of the Pilot Rig mirror machine, the first superconducting magnet facility to be used in plasma physics or fusion research; operation of the Superconducting Bumpy Torus magnet facility, first used to generate a toroidal magnetic field; steady state production of neutrons from DD reactions; studies of the direct conversion of plasma enthalpy to thrust by a direct fusion rocket via propellant addition and magnetic nozzles; power and propulsion system studies, including D(3)He power balance, neutron shielding, and refrigeration requirements; and development of large volume, high field superconducting and cryogenic magnet technology

Inertial fusion energy; L'energie de fusion inertielle

Energy Technology Data Exchange (ETDEWEB)

Decroisette, M.; Andre, M.; Bayer, C.; Juraszek, D. [CEA Bruyeres-le-Chatel, Dir. des Systemes d' Information (CEA/DIF), 91 (France); Le Garrec, B. [CEA Centre d' Etudes Scientifiques et Techniques d' Aquitaine, 33 - Le Barp (France); Deutsch, C. [Paris-11 Univ., 91 - Orsay (France); Migus, A. [Institut d' Optique Centre scientifique, 91 - Orsay (France)

2005-07-01

We first recall the scientific basis of inertial fusion and then describe a generic fusion reactor with the different components: the driver, the fusion chamber, the material treatment unit, the target factory and the turbines. We analyse the options proposed at the present time for the driver and for target irradiation scheme giving the state of art for each approach. We conclude by the presentation of LMJ (laser Megajoule) and NIF (national ignition facility) projects. These facilities aim to demonstrate the feasibility of laboratory DT ignition, first step toward Inertial Fusion Energy. (authors)

Operation of the cryogenic system for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Chronis, W.C.; Slack, D.S.

1987-01-01

The cryogenic system for the Mirror Fusion Test Facility (MFTF) at Lawrence Livermore National Laboratory (LLNL) was designed to cool the entire MFTF-B system from ambient to operating temperature in less than 10 days. The system was successfully operated in the recent plant and capital equipment (PACE) acceptance tests, and results from these tests helped us correct problem areas and improve the system

Sandia National Laboratories/New Mexico Facilities and Safety Information Document [NOTE: Volume II, Chapter 12

International Nuclear Information System (INIS)

March, F.; Guerrero, J.V.; Johns, W.H.; Schetnan, R.; Bayliss, L.S.; Kuzio, K.A.

1999-01-01

Operations in Tech Area IV commenced in 1980 with the construction of Buildings 980 and 981 and the Electron Beam Fusion Accelerator, which at the time was a major facility in SNL's Inertial Confinement Fusion Program. The Electron Beam Fusion Accelerator was a third-generation fusion accelerator that followed Proto I and Proto II, which were operated in Tech Area V. Another accelerator, the Particle Beam Fusion Accelerator I, was constructed in Tech Area IV because there was not enough room in Tech Area V, a highly restricted area that contains SNL's reactor facilities. In the early 1980s, more fusion-related facilities were constructed in Tech Area IV. Building 983 was built to house a fourth-generation fusion accelerator, the Particle Beam Fusion Accelerator II, now called Z Machine, and Buildings 960 and 961 were built to house office space, electrical and mechanical laboratories, and highbay space for pulsed power research and development. In the mid 1980s, Building 970 was constructed to house the Simulation Technology Laboratory. The main facility in the Simulation Technology Laboratory is the High-Energy Radiation Megavolt Electron Source (HERMES) III, a third-generation gamma ray accelerator that is used primarily for the simulation of gamma rays produced by nuclear weapons. The previous generations, HERMES I and HERMES II, had been located in Tech Area V. In the late 1980s, Proto II was moved from Tech Area V to the Simulation Technology Laboratory and modified to function as an x-ray simulation accelerator, and construction of Buildings 962 and 963 began. These buildings comprised the Strategic Defense Facility, which was initially intended to support the nation's Strategic Defense Initiative or ''Star Wars'' program. It was to house a variety of pulsed power-related facilities to conduct research in such areas as directed-energy weapons (electron beams, lasers, and microwaves) and an earth-to-orbit launcher. With the reduction of the Strategic Defense

Proposed design criteria for a fusion facility electrical ground system

International Nuclear Information System (INIS)

Armellino, C.A.

1983-01-01

Ground grid design considerations for a nuclear fusion reactor facility are no different than any other facility in that the basis for design must be safety first and foremost. Unlike a conventional industrial facility the available fault energy comes not only from the utility source and in-house rotating machinery, but also from energy storage capacitor banks, collapsing magnetic fields and D.C. transmission lines. It is not inconceivable for a fault condition occurrence where all available energy can be discharged. The ground grid must adequately shunt this sudden energy discharge in a way that personnel will not be exposed by step and/or touch to hazardous energy levels that are in excess of maximum tolerable levels for humans. Fault energy discharge rate is a function of the ground grid surge impedance characteristic. Closed loop paths must be avoided in the ground grid design so that during energy discharge no stray magnetic fields or large voltage potentials between remote points can be created by circulating currents. Single point connection of equipment to the ground grid will afford protection to personnel and sensitive equipment by reducing the probability of circulating currents. The overall ground grid system design is best illustrated as a wagon wheel concept with the fusion machine at the center. Radial branches or spokes reach out to the perimeter limits designated by step-and-touch high risk areas based on soil resistivity criteria considerations. Conventional methods for the design of a ground grid with all of its radial branches are still pertinent. The center of the grid could include a deep well single ground rod element the length of which is at least equivalent to the radius of an imaginary sphere that enshrouds the immediate machine area. Special facilities such as screen rooms or other shielded areas are part of the ground grid system by way of connection to radial branches

Fusion power research and development program. Volume IV. 5-year program, budget and milestone summaries

International Nuclear Information System (INIS)

1976-07-01

Budget data are given for each of the tokamak systems, mirror systems, and high density plasma systems for the years 1976 through 1982. All major facilities currently under ERDA contract are included. In addition, budget data are given for the development and technology program consisting of the following; (1) magnetic systems, (2) plasma engineering, (3) fusion reactor materials, (4) fusion systems engineering, (5) environment and safety, and (6) applied plasma physics

FINESSE: study of the issues, experiments and facilities for fusion nuclear technology research and development. Interim report. Volume II

International Nuclear Information System (INIS)

Abdou, M.

1984-10-01

The Nuclear Fusion Issues chapter contains a comprehensive list of engineering issues for fusion reactor nuclear components. The list explicitly defines the uncertainties associated with the engineering option of a fusion reactor and addresses the potential consequences resulting from each issue. The next chapter identifies the fusion nuclear technology testing needs up to the engineering demonstration stage

Shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

Carter, L.L.; Mann, F.M.; Morford, R.J.; Wilcox, A.D.; Johnson, D.L.; Huang, S.T.

1983-03-01

The shield design for the Fusion Materials Irradiation Test facility is based upon one-, two- and three-dimensional transport calculations with experimental measurements utilized to refine the nuclear data including the neutron cross sections from 20 to 50 MeV and the gamma ray and neutron source terms. The high energy neutrons and deuterons produce activation products from the numerous reactions that are kinematically allowed. The analyses for both beam-on and beam-off (from the activation products) conditions have required extensive nuclear data libraries and the utilization of Monte Carlo, discrete ordinates, point kernel and auxiliary computer codes

Calculation of damage function of Al{sub 2}O{sub 3} in irradiation facilities for fusion reactor applications

Energy Technology Data Exchange (ETDEWEB)

Mota, F., E-mail: fernando.mota@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Ortiz, C.J., E-mail: christophe.ortiz@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Vila, R., E-mail: rafael.vila@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Casal, N., E-mail: natalia.casal@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); García, A., E-mail: angela.garcia@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain); Ibarra, A., E-mail: Angel.ibarra@ciemat.es [Laboratorio Nacional de Fusión por Confinamiento Magnético – CIEMAT, 28040 Madrid (Spain)

2013-11-15

A rigorous material testing program is essential for the development of the nuclear fusion world program. In particular, it is very important to predict the generation of the displacement damage in materials, because the irradiation intensity expected in fusion conditions is such that the performance of materials and components under these extreme conditions is unknown. To study the damage produced by neutrons in materials of interest for fusion, a specific computational methodology was developed. Neutron fluxes expected in different irradiation facilities (International Fusion Materials Irradiation Facility [IFMIF] and DEMO-HCLL) and in different irradiation spots were obtained with particles transport codes (McDeLicious, MCNP). The energy differential cross sections of primary knock-on atoms were calculated using the NJOY code. Resulting data were input into the Monte Carlo code MARLOWE to calculate the corresponding displacements (i.e., interstitials (I) and vacancies (V)). However, the number of Frenkel pairs created during irradiation strongly depends on the recombination radius between interstitials and vacancies. This parameter corresponds to the minimum distance below which instantaneous recombination occurs. Mainly, the influence of such parameter on the damage function in Al{sub 2}O{sub 3} was assessed in this report. The displacements per atom values calculated as a function of the recombination radius considered are compared to experimental data to determine the most appropriate capture radius. In addition, the damage function and damage dose generated at different experimental irradiation facilities are compared with those expected in DEMO. The conclusion is that both IFMIF and TechnoFusión (future triple beam ion accelerator to emulate fusion neutron irradiation effects in materials) facilities are suited to perform relevant irradiation experiments for the design of DEMO.

Experimental fusion power reactor conceptual design study. Final report. Volume III

International Nuclear Information System (INIS)

Baker, C.C.

1976-12-01

This document is the final report which describes the work carried out by General Atomic Company for the Electric Power Research Institute on a conceptual design study of a fusion experimental power reactor (EPR) and an overall EPR facility. The primary objective of the two-year program was to develop a conceptual design of an EPR that operates at ignition and produces continuous net power. A conceptual design was developed for a Doublet configuration based on indications that a noncircular tokamak offers the best potential of achieving a sufficiently high effective fuel containment to provide a viable reactor concept at reasonable cost. Other objectives included the development of a planning cost estimate and schedule for the plant and the identification of critical R and D programs required to support the physics development and engineering and construction of the EPR. This volume contains the following appendices: (1) tradeoff code analysis, (2) residual mode transport, (3) blanket/first wall design evaluations, (4) shielding design evaluation, (5) toroidal coil design evaluation, (6) E-coil design evaluation, (7) F-coil design evaluation, (8) plasma recycle system design evaluation, (9) primary coolant purification design evaluation, (10) power supply system design evaluation, (11) number of coolant loops, (12) power conversion system design evaluation, and (13) maintenance methods evaluation

Interactive dual-volume rendering visualization with real-time fusion and transfer function enhancement

Science.gov (United States)

Macready, Hugh; Kim, Jinman; Feng, David; Cai, Weidong

2006-03-01

Dual-modality imaging scanners combining functional PET and anatomical CT constitute a challenge in volumetric visualization that can be limited by the high computational demand and expense. This study aims at providing physicians with multi-dimensional visualization tools, in order to navigate and manipulate the data running on a consumer PC. We have maximized the utilization of pixel-shader architecture of the low-cost graphic hardware and the texture-based volume rendering to provide visualization tools with high degree of interactivity. All the software was developed using OpenGL and Silicon Graphics Inc. Volumizer, tested on a Pentium mobile CPU on a PC notebook with 64M graphic memory. We render the individual modalities separately, and performing real-time per-voxel fusion. We designed a novel "alpha-spike" transfer function to interactively identify structure of interest from volume rendering of PET/CT. This works by assigning a non-linear opacity to the voxels, thus, allowing the physician to selectively eliminate or reveal information from the PET/CT volumes. As the PET and CT are rendered independently, manipulations can be applied to individual volumes, for instance, the application of transfer function to CT to reveal the lung boundary while adjusting the fusion ration between the CT and PET to enhance the contrast of a tumour region, with the resultant manipulated data sets fused together in real-time as the adjustments are made. In addition to conventional navigation and manipulation tools, such as scaling, LUT, volume slicing, and others, our strategy permits efficient visualization of PET/CT volume rendering which can potentially aid in interpretation and diagnosis.

Inertial fusion energy

International Nuclear Information System (INIS)

Decroisette, M.; Andre, M.; Bayer, C.; Juraszek, D.; Le Garrec, B.; Deutsch, C.; Migus, A.

2005-01-01

We first recall the scientific basis of inertial fusion and then describe a generic fusion reactor with the different components: the driver, the fusion chamber, the material treatment unit, the target factory and the turbines. We analyse the options proposed at the present time for the driver and for target irradiation scheme giving the state of art for each approach. We conclude by the presentation of LMJ (laser Megajoule) and NIF (national ignition facility) projects. These facilities aim to demonstrate the feasibility of laboratory DT ignition, first step toward Inertial Fusion Energy. (authors)

A spallation-based irradiation test facility for fusion and future fission materials

CERN Document Server

Samec, K; Kadi, Y; Luis, R; Romanets, Y; Behzad, M; Aleksan, R; Bousson, S

2014-01-01

The EU’s FP7 TIARA program for developing accelerator-based facilities has recently demonstrated the unique capabilities of a compact and powerful spallation source for irradiating advanced nuclear materials. The spectrum and intensity of the neutron flux produced in the proposed facility fulfils the requirements of the DEMO fusion reactor for ITER, ADS reactors and also Gen III / IV reactors. Test conditions can be modulated, covering temperature from 400 to 550°C, liquid metal corrosion, cyclical or static stress up to 500 MPa and neutron/proton irradiation damage of up to 25 DPA per annum. The entire “TMIF” facility fits inside a cube 2 metres on a side, and is dimensioned for an accelerator beam power of 100 kW, thus reducing costs and offering great versatility and flexibility.

System of the advanced volume reduction facilities for LLW at JAERI

International Nuclear Information System (INIS)

Higuchi, Hidekazu; Monma, Toshiyuki; Nakashio, Nobuyuki; Kozawa, Kazushige; Touhei, Toshio; Sudou, Tomoyuki; Mitsuda, Motoyuki; Kurosawa, Shigenobu; Henmi, Kou; Ishikawa, Joji; Kato, Mitsugu; Sato, Motoaki

2005-01-01

The Japan Atomic Energy Research Institute (JAERI) constructed the Advanced Volume Reduction Facilities (AVRF), in which volume reduction techniques are applied and achieved high volume reduction ratio, homogenization and stabilization by means of melting or super compaction processes for low level solid wastes. It will be able to produce waste packages for final disposal and to reduce the amount of the wastes by operating the AVRF. The AVRF consist of the Waste Size Reduction and Storage Facilities (WSRSF) and the Waste Volume Reduction Facilities (WVRF); the former have cutting installations for large size wastes and the latter have melting units and a super compactor. Cutting installations in the WSRSF have been operating since June 1999. Radioactive wastes treated so far amount to 600 m 3 and the volume reduction ratio is from 1/2 to 1/3. The WVRF have been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation with radioactive wastes will start in FY2005. (author)

Contributions of the National Ignition Facility to the development of Inertial Fusion Energy

International Nuclear Information System (INIS)

Tobin, M.; Logan, G.; Diaz De La Rubia, T.; Schrock, V.; Schultz, K.; Tokheim, R.; Abdou, M.; Bangerter, R.

1994-06-01

The Department of Energy is proposing to construct the National Ignition Facility (NIF) to embark on a program to achieve ignition and modest gain in the laboratory early in the next century. The NIF will use a ≥ 1.8-MJ, 0.35-mm laser with 192 independent beams, a fifty-fold increase over the energy of the Nova laser. System performance analyses suggest yields as great as 20 MJ may be achievable. The benefits of a micro-fusion capability in the laboratory include: essential contributions to defense programs, resolution of important Inertial Fusion Energy issues, and unparalleled conditions of energy density for basic science and technology research. We have begun to consider the role the National Ignition Facility will fill in the development of Inertial Fusion Energy. While the achievement of ignition and gain speaks for itself in terms of its impact on developing IFE, we believe there are areas of IFE development such as fusion power technology, IFE target design and fabrication, and understanding chamber dynamics, that would significantly benefit from NIF experiments. In the area of IFE target physics, ion targets will be designed using the NIF laser, and feasibility of high gain targets will be confirmed. Target chamber dynamics experiments will benefit from x-ray and debris energies that mimic in-IFE-chamber conditions. Fusion power technology will benefit from using single-shot neutron yields to measure spatial distribution of neutron heating, activation, and tritium breeding in relevant materials. IFE target systems will benefit from evaluating low-cost target fabrication techniques by testing such targets on NIF. Additionally, we believe it is feasible to inject up to four targets and engage them with the NIF laser by triggering the beams in groups of ∼50 separated in time by ∼0.1 s. Sub-ignition neutron yields would allow an indication of symmetry achieved in such proof-of-principle rep-rate experiments

A remote monitoring system of environmental electromagnetic field in magnetic confinement fusion test facilities

International Nuclear Information System (INIS)

Tanaka, Masahiro; Uda, Tatsuhiko; Takami, Shigeyuki; Wang, Jianqing; Fujiwara, Osamu

2010-01-01

A remote, continuous environmental electromagnetic field monitoring system for use in magnetic confinement fusion test facilities is developed. Using this system, both the static magnetic field and the high frequency electromagnetic field could be measured. The required frequency range of the measurement system is from 25 to 100 MHz for the ICRF (Ion Cyclotron Range of Frequencies) heating system. The outputs from the measurement instruments are measured simultaneously by custom-built software using a laptop-type personal computer connected to a local area network. In this way, the electromagnetic field strength could be monitored from a control room located about 200 m from the fusion device building. Examples of measurement data from the vicinity of a high-frequency generator and amplifier and the leakage static magnetic field from a fusion test device are presented. (author)

International fusion materials irradiation facility and neutronic calculations for its test modules

International Nuclear Information System (INIS)

Sokcic-Kostic, M.

1997-01-01

The International Fusion Material Irradiation Facility (IFMIF) is a projected high intensity neutron source for material testing. Neutron transport calculations for the IFMIF project are performed for variety of here explained reasons. The results of MCNP neutronic calculations for IFMIF test modules with NaK and He cooled high flux test cells are presented in this paper. (author). 3 refs., 2 figs., 3 tabs

Early Career. Harnessing nanotechnology for fusion plasma-material interface research in an in-situ particle-surface interaction facility

Energy Technology Data Exchange (ETDEWEB)

Allain, Jean Paul [Univ. of Illinois, Champaign, IL (United States)

2014-08-08

This project consisted of fundamental and applied research of advanced in-situ particle-beam interactions with surfaces/interfaces to discover novel materials able to tolerate intense conditions at the plasma-material interface (PMI) in future fusion burning plasma devices. The project established a novel facility that is capable of not only characterizing new fusion nanomaterials but, more importantly probing and manipulating materials at the nanoscale while performing subsequent single-effect in-situ testing of their performance under simulated environments in fusion PMI.

Tools for remote collaboration on the DIII-D national fusion facility

International Nuclear Information System (INIS)

McHarg, B.B. Jr.; Greenwood, D.

1999-01-01

The DIII-D national fusion facility, a tokamak experiment funded by the US Department of Energy and operated by General Atomics (GA), is an international resource for plasma physics and fusion energy science research. This facility has a long history of collaborations with scientists from a wide variety of laboratories and universities from around the world. That collaboration has mostly been conducted by travel to and participation at the DIII-D site. Many new developments in the computing and technology fields are now facilitating collaboration from remote sites, thus reducing some of the needs to travel to the experiment. Some of these developments include higher speed wide area networks, powerful workstations connected within a distributed computing environment, network based audio/video capabilities, and the use of the world wide web. As the number of collaborators increases, the need for remote tools become important options to efficiently utilize the DIII-D facility. In the last two years a joint study by GA, Princeton Plasma Physics Laboratory (PPPL), Lawrence Livermore National Laboratory (LLNL), and Oak Ridge National Laboratory (ORNL) has introduced remote collaboration tools into the DIII-D environment and studied their effectiveness. These tools have included the use of audio/video for communication from the DIII-D control room, the broadcast of meetings, use of inter-process communication software to post events to the network during a tokamak shot, the creation of a DCE (distributed computing environment) cell for creating a common collaboratory environment, distributed use of computer cycles, remote data access, and remote display of results. This study also included sociological studies of how scientists in this environment work together as well as apart. (orig.)

Complex workplace radiation fields at European high-energy accelerators and thermonuclear fusion facilities

CERN Document Server

Bilski, P; D'Errico, F; Esposito, A; Fehrenbacher, G; Fernàndez, F; Fuchs, A; Golnik, N; Lacoste, V; Leuschner, A; Sandri, S; Silari, M; Spurny, F; Wiegel, B; Wright, P

2006-01-01

This report outlines the research needs and research activities within Europe to develop new and improved methods and techniques for the characterization of complex radiation fields at workplaces around high-energy accelerators and the next generation of thermonuclear fusion facilities under the auspices of the COordinated Network for RAdiation Dosimetry (CONRAD) project funded by the European Commission.

Manufacturing cost study on the ion sources for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

A study of the cost of manufacturing 48 ion sources for the Mirror Fusion Test Facility is described. The estimate is built up from individual part costs and assembly operation times for the 80 kV prototype source constructed by LLL and described by LLL drawings furnished during December 1978. Recommendations for cost reduction are made

How to improve the irradiation conditions for the International Fusion Materials Irradiation Facility

CERN Document Server

Daum, E

2000-01-01

The accelerator-based intense D-Li neutron source International Fusion Materials Irradiation Facility (IFMIF) provides very suitable irradiation conditions for fusion materials development with the attractive option of accelerated irradiations. Investigations show that a neutron moderator made of tungsten and placed in the IFMIF test cell can further improve the irradiation conditions. The moderator softens the IFMIF neutron spectrum by enhancing the fraction of low energy neutrons. For displacement damage, the ratio of point defects to cascades is more DEMO relevant and for tritium production in Li-based breeding ceramic materials it leads to a preferred production via the sup 6 Li(n,t) sup 4 He channel as it occurs in a DEMO breeding blanket.

Trial operation of the advanced volume reduction facilities for LLW at JAEA

International Nuclear Information System (INIS)

Nakashio, Nobuyuki; Higuchi, Hidekazu; Momma, Toshiyuki; Kozawa, Kazushige; Touhei, Toshio; Sudou, Tomoyuki; Mitsuda, Motoyuki; Kurosawa, Shigenobu; Hemmi, Kou; Ishikawa, Joji; Kato, Mitsugu; Sato, Motoaki

2007-01-01

The Japan Atomic Energy Agency (JAEA) constructed the Advanced Volume Reduction Facilities (AVRF), in which volume reduction techniques are applied and achieved high volume reduction ratio, homogenization and stabilization by means of melting or super compaction processes for low level radioactive solid wastes. It will be able to produce waste packages for final disposal and to reduce the volume of stored wastes by operating the AVRF. The AVRF consist of the Waste Size Reduction and Storage Facilities (WSRSF) and the Waste Volume Reduction Facilities (WVRF); the former has cutting installations for large size wastes and the latter has melting units and a super compactor. Cutting installations in the WSRSF have been operating since July 1999. Radioactive wastes treated so far amount to 750 m 3 and the volume reduction ratio is from 1.7 to 3.7. The WVRF has been operating with non-radioactive wastes since February 2003 for the training and the homogeneity investigation in the melting processes. The operation of the pretreatment system in the WVRF with radioactive wastes has partly started in FY2005. (author)

An overview of safety and environmental considerations in the selection of materials for fusion facilities

International Nuclear Information System (INIS)

Petti, D.A.; Piet, S.J.; Seki, Y.

1996-01-01

Safety and environmental considerations can play a large role in the selection of fusion materials. In this paper, we review the attributes of different structural, plasma facing, and breeding materials from a safety perspective and discuss some generic waste management issues as they relate to fusion materials in general. Specific safety concerns exist for each material that must be dealt with in fusion facility design. Low activation materials offer inherent safety benefits compared with conventional materials, but more work is needed before these materials have the requisite certified databases. In the interim, the international thermonuclear experimental reactor (ITER) has selected more conventional materials and is showing that the safety concerns with these materials can be addressed by proper attention to design. In the area of waste management disposal criteria differ by country. However, the criteria are all very strict making disposal of fusion components difficult. As a result, recycling has gained increasing attention. (orig.)

OSIRIS and SOMBRERO Inertial Fusion Power Plant Designs, Volume 2: Designs, Assessments, and Comparisons

Energy Technology Data Exchange (ETDEWEB)

Meier, W. R.; Bieri, R. L.; Monsler, M. J.; Hendricks, C. D.; Laybourne, P.; Shillito, K. R.

1992-03-01

This is a comprehensive design study of two Inertial Fusion Energy (IFE) electric power plants. Conceptual designs are presented for a fusion reactor (called Osiris) using an induction-linac heavy-ion beam driver, and another (called SOMBRERO) using a KrF laser driver. The designs covered all aspects of IFE power plants, including the chambers, heat transport and power conversion systems, balance-of-plant facilities, target fabrication, target injection and tracking, as well as the heavy-ion and KrF drivers. The point designs were assessed and compared in terms of their environmental & safety aspects, reliability and availability, economics, and technology development needs.

Fusion technology 1992

International Nuclear Information System (INIS)

Ferro, C.; Gasparatto, M.; Knoepfel, H.

1993-01-01

The aim of the biennial series of symposia on the title subject, organized by the European Fusion Laboratories, is the exchange of information on the design, construction and operation of fusion experiments and on the technology being developed for the next step devices and fusion reactors. The coverage of the volume includes the technological aspects of fusion reactors in relation to new developments, this forming a guideline for the definition of future work. These proceedings comprise three volumes and contain both the invited lectures and contributed papers presented at the symposium which was attended by 569 participants from around the globe. The 343 papers, including 12 invited papers, characterize the increasing interest of industry in the fusion programme, giving a broad and current overview on the progress and trends fusion technology is experiencing now, as well as indicating the future for fusion devices

Seismic analysis of the mirror fusion test facility shielding vault

International Nuclear Information System (INIS)

Gabrielsen, B.L.; Tsai, K.

1981-04-01

This report presents a seismic analysis of the vault in Building 431 at Lawrence Livermore National Laboratory which houses the mirror Fusion Test Facility. The shielding vault structure is approximately 120 ft long by 80 ft wide and is constructed of concrete blocks approximately 7 x 7 x 7 ft. The north and south walls are approximately 53 ft high and the east wall is approximately 29 ft high. These walls are supported on a monolithic concrete foundation that surrounds a 21-ft deep open pit. Since the 53-ft walls appeared to present the greatest seismic problem they were the first investigated

World Energy Data System (WENDS). Volume X. Nuclear facility profiles, PO--ZA

International Nuclear Information System (INIS)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile

World Energy Data System (WENDS). Volume VIII. Nuclear facility profiles, CO--HU

International Nuclear Information System (INIS)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile

World Energy Data System (WENDS). Volume IX. Nuclear facility profiles, IN--PL

International Nuclear Information System (INIS)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile

World Energy Data System (WENDS). Volume VII. Nuclear facility profiles, AG--CH

International Nuclear Information System (INIS)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile

IFMIF (International Fusion Materials Irradiation Facility) conceptual design activity reduced cost report

International Nuclear Information System (INIS)

2000-02-01

This report describes the results of a preliminary reevaluation of the design and cost of the International Fusion Materials Irradiation Facility (IFMIF) Project in response to the request from the 28th FPCC meeting in January 1999. Two major ideas have been considered: 1) reduction of the total construction cost through elimination of the previously planned facility upgrade and 2) a facility deployment in 3 stages with capabilities for limited experiments in the first stage. As a result, the size and complexity of the facility could be significantly reduced, leading to substantial cost savings. In addition to these two ideas, this study also included a critical review of the original CDA specification with the objective of elimination of nonessential items. For example, the number of lithium targets was reduced from two to one. As a result of these changes in addition to the elimination of the upgrade, the total cost estimate was very substantially reduced from 797.2 MICF to 487.8 MICF, where 1 MICF = 1 Million of the IFMIF Conversion Units (approximately $1M US January, 1996). (author)

Development of fusion safety standards

International Nuclear Information System (INIS)

Longhurst, G.R.; Petti, D.A.; Dinneen, G.A.; Herring, J.S.; DeLooper, J.; Levine, J.D.; Gouge, M.J.

1996-01-01

Two new U.S. Department of Energy (DOE) standards have been prepared to assist in the design and regulation of magnetic fusion facilities. They are DOE-STD-6002-96, 'Safety of Magnetic Fusion Facilities - Requirements,' and DOE-STD-6003-96 'Safety of Magnetic Fusion Facilities - Guidance.' The first standard sets forth requirements, mostly based on the Code of Federal Regulations, deemed necessary for the safe design and operation of fusion facilities and a set of safety principles to use in the design. The second standard provides guidance on how to meet the requirements identified in DOE-STD-6002-96. It is written specifically for a facility such as the International Thermonuclear Experimental Reactor (ITER) in the DOE regulatory environment. As technical standards, they are applicable only to the extent that compliance with these standards is included in the contracts of the developers. 7 refs., 1 fig

Laser-Plasma Interactions on NIKE and the Fusion Test Facility

Science.gov (United States)

Phillips, Lee; Weaver, James

2008-11-01

Recent proposed designs for a Fusion Test Facility (FTF) (Obenchain et al., Phys. Plasmas 13 056320 (2006)) for direct-drive ICF targets for energy applications involve high implosion velocities combined with higher laser irradiances. The use of high irradiances increases the likelihood of deleterious laser plasma instabilities (LPI) but the proposed use of a 248 nm KrF laser to drive these targets is expected to minimize the LPI risk. We examine, using simulation results from NRL's FAST hydrocode, the proposed operational regimes of the FTF in relation to the thresholds for the SRS, SBS, and 2-plasmon instabilities. Simulations are also used to help design and interpret ongoing experiments being conducted at NRL's NIKE facility for the purpose of generating and studying LPI. Target geometries and laser pulseshapes were devised in order to create plasma conditions with long scalelengths and low electron temperatures that allow the growth of parametric instabilities. These simulations include the effects of finite beam angles through the use of raytracing.

Recent progress in ignition fusion research on the National Ignition Facility

International Nuclear Information System (INIS)

Leeper, Ramon J.

2011-01-01

This paper will review the ignition fusion research program that is currently being carried out on the National Ignition Facility (NIF) located at Lawrence Livermore National Laboratory. This work is being conducted under the auspices of the National Ignition Campaign (NIC) that is a broad collaboration of national laboratories and universities that together have developed a detailed research plan whose goal is ignition in the laboratory. The paper will begin with a description of the NIF facility and associated experimental facilities. The paper will then focus on the ignition target and hohlraum designs that will be tested in the first ignition attempts on NIF. The next topic to be introduced will be a description of the diagnostic suite that has been developed for the initial ignition experiments on NIF. The paper will then describe the experimental results that were obtained in experiments conducted during the fall of 2009 on NIF. Finally, the paper will end with a description of the detailed experimental plans that have been developed for the first ignition campaign that will begin later this year. (author)

ORNL facilities for testing first-wall components

International Nuclear Information System (INIS)

Tsai, C.C.; Becraft, W.R.; Gardner, W.L.; Haselton, H.H.; Hoffman, D.J.; Menon, M.M.; Stirling, W.L.

1985-01-01

Future long-impulse magnetic fusion devices will have operating characteristics similar to those described in the design studies of the Tokamak Fusion Core Experiment (TFCX), the Fusion Engineering Device (FED), and the International Tokamak Reactor (INTOR). Their first-wall components (pumped limiters, divertor plates, and rf waveguide launchers with Faraday shields) will be subjected to intense bombardment by energetic particles exhausted from the plasma, including fusion products. These particles are expected to have particle energies of approx.100 eV, particle fluxes of approx.10 18 cm -2 .s -1 , and heat fluxes of approx.1 kW/cm 2 CW to approx.100 kW/cm 2 transient. No components are available to simultaneously handle these particle and heat fluxes, survive the resulting sputtering erosion, and remove exhaust gas without degrading plasma quality. Critical issues for research and development of first-wall components have been identified in the INTOR Activity. Test facilities are needed to qualify candidate materials and develop components. At Oak Ridge National Laboratory (ORNL), existing neutral beam and wave heating test facilities can be modified to simulate first-wall environments with heat fluxes up to 30 kW/cm 2 , particle fluxes of approx.10 18 cm -2 .s -1 , and pulse lengths up to 30 s, within test volumes up to approx.100 L. The characteristics of these test facilities are described, with particular attention to the areas of particle flux, heat flux, particle energy, pulse length, and duty cycle, and the potential applications of these facilities for first-wall component development are discussed

Analysis of the tritium-water (T-H2O) system for a fusion material test facility

International Nuclear Information System (INIS)

Hassanein, A.; Smith, D.L.; Sze, D.K.; Reed, C.B.

1992-04-01

The need for a high flux, high energy neutron test facility to evaluate performance of fusion reactor materials is urgent. An accelerator based D-Li source is generally accepted as the most reasonable approach to a high flux neutron source in the near future. The idea is to bombard a high energy (35 MeV) deuteron beam into a lithium target to produce high energy neutrons to simulate the fusion environment. More recently it was proposed to use a 21 MeV triton beam incident on a water jet target to produce the required neutron source for testing and simulating fusion material environments. The advantages of such a system are discussed. Major concerns regarding the feasibility of this system are also highlighted

Fusion technology development: first wall/blanket system and component testing in existing nuclear facilities

International Nuclear Information System (INIS)

Hsu, P.Y.S.; Bohn, T.S.; Deis, G.A.; Judd, J.L.; Longhurst, G.R.; Miller, L.G.; Millsap, D.A.; Scott, A.J.; Wessol, D.E.

1980-12-01

A novel concept to produce a reasonable simulation of a fusion first wall/blanket test environment employing an existing nuclear facility, the Engineering Test Reactor at the Idaho National Engineering Laboratory, is presented. Preliminary results show that an asymmetric, nuclear test environment with surface and volumetric heating rates similar to those expected in a fusion first wall/blanket or divertor chamber surface appears feasible. The proposed concept takes advantage of nuclear reactions within the annulus of an existing test space (15 cm in diameter and approximately 100 cm high) to provide an energy flux to the surface of a test module. The principal reaction considered involves 3 He in the annulus as follows: n + 3 He → p + t + 0.75 MeV. Bulk heating in the test module is accomplished by neutron thermalization, gamma heating, and absorption reactions involving 6 Li in the blanket breeding region. The concept can be extended to modified core configurations that will accommodate test modules of different sizes and types. It makes possible development testing of first wall/blanket systems and other fusion components on a scale and in ways not otherwise available until actual high-power fusion reactors are built

First downscattered neutron images from Inertial Confinement Fusion experiments at the National Ignition Facility

Directory of Open Access Journals (Sweden)

Guler Nevzat

2013-11-01

Full Text Available Inertial Confinement Fusion experiments at the National Ignition Facility (NIF are designed to understand and test the basic principles of self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT filled cryogenic plastic (CH capsules. The experimental campaign is ongoing to tune the implosions and characterize the burning plasma conditions. Nuclear diagnostics play an important role in measuring the characteristics of these burning plasmas, providing feedback to improve the implosion dynamics. The Neutron Imaging (NI diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by collecting images at two different energy bands for primary (13–15 MeV and downscattered (10–12 MeV neutrons. From these distributions, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. The first downscattered neutron images from imploding ICF capsules are shown in this paper.

First downscattered neutron images from Inertial Confinement Fusion experiments at the National Ignition Facility

Science.gov (United States)

Guler, Nevzat; Aragonez, Robert J.; Archuleta, Thomas N.; Batha, Steven H.; Clark, David D.; Clark, Deborah J.; Danly, Chris R.; Day, Robert D.; Fatherley, Valerie E.; Finch, Joshua P.; Gallegos, Robert A.; Garcia, Felix P.; Grim, Gary; Hsu, Albert H.; Jaramillo, Steven A.; Loomis, Eric N.; Mares, Danielle; Martinson, Drew D.; Merrill, Frank E.; Morgan, George L.; Munson, Carter; Murphy, Thomas J.; Oertel, John A.; Polk, Paul J.; Schmidt, Derek W.; Tregillis, Ian L.; Valdez, Adelaida C.; Volegov, Petr L.; Wang, Tai-Sen F.; Wilde, Carl H.; Wilke, Mark D.; Wilson, Douglas C.; Atkinson, Dennis P.; Bower, Dan E.; Drury, Owen B.; Dzenitis, John M.; Felker, Brian; Fittinghoff, David N.; Frank, Matthias; Liddick, Sean N.; Moran, Michael J.; Roberson, George P.; Weiss, Paul; Buckles, Robert A.; Cradick, Jerry R.; Kaufman, Morris I.; Lutz, Steve S.; Malone, Robert M.; Traille, Albert

2013-11-01

Inertial Confinement Fusion experiments at the National Ignition Facility (NIF) are designed to understand and test the basic principles of self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT) filled cryogenic plastic (CH) capsules. The experimental campaign is ongoing to tune the implosions and characterize the burning plasma conditions. Nuclear diagnostics play an important role in measuring the characteristics of these burning plasmas, providing feedback to improve the implosion dynamics. The Neutron Imaging (NI) diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by collecting images at two different energy bands for primary (13-15 MeV) and downscattered (10-12 MeV) neutrons. From these distributions, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. The first downscattered neutron images from imploding ICF capsules are shown in this paper.

H-1NF: Australian national fusion plasma research facility

International Nuclear Information System (INIS)

Blackwell, B.D.; Borg, G.G.; Dewar, R.L.; Howard, J.; Gardner, H.J.; Rudakov, D.L.; Sharp, L.E.; Shats, M.G.; Warr, G.B.

1997-01-01

The H-1 heliac is a helical axis stellarator of moderate size and novel, flexible configuration. Since commissioning, H-1 has operated in quasi-continuous mode at low magnetic field. For higher fields ≤1T an ECRH heating system (28GHz, 200kW) has been installed under a collaborative agreement between ANU and NIFS. H-1 has recently been promoted to national facility status (H-1NF), which will include upgrades of the rf and ech heating systems to megawatt powers, and power supply and diagnostic and data system enhancements. This facilitates collaborative research locally (through the Australian Fusion Research Group consortium) and internationally. Results of a number of basic experiments in quasi-continuous mode are presented. (author)

Accelerator conceptual design of the international fusion materials irradiation facility

International Nuclear Information System (INIS)

Sugimoto, M.; Kinsho, M.; Teplyakov, V.; Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J.; Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K.; Miyahara, A.; Olivier, M.; Piechowiak, E.; Tanabe, Y.

1998-01-01

The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.)

Application of the INS facility as a high-flux benchmark for neutron dosimetry and for radiation damage studies in D--T fusion spectra

International Nuclear Information System (INIS)

Dierckx, R.; Emigh, C.R.

1977-01-01

An Intense Neutron Source facility (INS), is presently under construction at the Los Alamos Scientific Laboratory. This facility is being built by the Energy Research and Development Administration for the radiation damage program in magnetic fusion energy. The facility will contain two D-T neutron sources, both producing about 10 15 primary 14-MeV neutrons per second on a continuous basis. One source will be used to produce a ''pure'' 14-MeV spectrum while the other will be surrounded by a multiplying blanket converter to produce a fusion-like spectrum with a total of about 10 16 neutrons per second

Radioactive waste package assay facility. Volume 1. Application of assay technology

International Nuclear Information System (INIS)

Findlay, D.J.S.; Green, T.H.; Molesworth, T.V.; Staniforth, D.; Strachan, N.R.; Rogers, J.D.; Wise, M.O.; Forrest, K.R.

1992-01-01

This report, in three volumes, covers the work carried out by Taylor Woodrow Construction Ltd., and two major sub-contractors: Harwell Laboratory (AEA Technology) and Siemens Plessey Controls Ltd., on the development of a radioactive waste package assay facility, for cemented 500 litre intermediate level waste drums. In volume 1, the reasons for assay are considered together with the various techniques that can be used, and the information that can be obtained. The practical problems associated with the use of the various techniques in an integrated assay facility are identified, and the key parameters defined. Engineering and operational features are examined and provisional designs proposed for facilities at three throughput levels: 15,000, 750 and 30 drums per year respectively. The capital and operating costs for such facilities have been estimated. A number of recommendations are made for further work. 16 refs., 14 figs., 13 tabs

Radioactive waste package assay facility. Volume 3. Data processing

International Nuclear Information System (INIS)

Creamer, S.C.; Lalies, A.A.; Wise, M.O.

1992-01-01

This report, in three volumes, covers the work carried out by Taylor Woodrow Construction Ltd, and two major sub-contractors: Harwell Laboratory (AEA Technology) and Siemens Plessey Controls Ltd, on the development of a radioactive waste package assay facility, for cemented 500 litre intermediate level waste drums. Volume 3, describes the work carried out by Siemens Plessey Controls Ltd on the data-processing aspects of an integrated waste assay facility. It introduces the need for a mathematical model of the assay process and develops a deterministic model which could be tested using Harwell experimental data. Relevant nuclear reactions are identified. Full implementation of the model was not possible within the scope of the Harwell experimental work, although calculations suggested that the model behaved as predicted by theory. 34 figs., 52 refs., 5 tabs

A spheromak ignition experiment reusing Mirror Fusion Test Facility (MFTF) equipment

International Nuclear Information System (INIS)

Fowler, T.K.

1993-01-01

Based on available experimental results and theory, a scenario is presented to achieve ohmic ignition in a spheromak by slow (∼ 10 sec.) helicity injection using power from the Mirror Fusion Test Facility (MFTF) substation. Some of the other parts needed (vacuum vessel, coils, power supplies, pumps, shielded building space) might also be obtained from MFTF or other salvage, as well as some components needed for intermediate experiments for additional verification of the concept (especially confinement scaling). The proposed ignition experiment would serve as proof-of-principle for the spheromak DT fusion reactor design published by Hagenson and Krakowski, with a nuclear island cost about ten times less than a tokamak of comparable power. Designs at even higher power density and lower cost might be possible using Christofilos' concept of a liquid lithium blanket. Since all structures would be protected from neutrons by the lithium blanket and the tritium inventory can be reduced by continuous removal from the liquid blanket, environmental and safety characteristics appear to be favorable

JSME construction standard for superconducting magnets of fusion facilities. Toward the construction of ITER

International Nuclear Information System (INIS)

Nakasone, Yuji; Takahashi, Yukio; Sato, Kazuyoshi; Nishimura, Arata; Suzuki, Tetsuya; Irie, Hirosada; Nakahira, Masataka

2009-01-01

The present paper describes the general view of the construction standard, which the Japan Society of Mechanical Engineers (JSME) has recently set up and published, for superconducting magnet structures to be used in nuclear fusion facilities. The present target of the standard is tokamak-type fusion energy facilities, especially the International Thermonuclear Experimental Reactor called ITER for short. The standard contains rules for structural materials including cryogenic materials, structural design considering magnetic forces, manufacture including welding and installation, nondestructive testing, pressure proof tests and leak tests of toroidal field magnet structures. The standard covers requirements for structural integrity, deformation control, and leak tightness of all the components of the superconducting magnets and their supports except for superconducting strands and electrical insulators. The standard does not cover deterioration, which may occur in service as a result of corrosion, radiation effects, or instability of material. The standard consists of seven articles and twelve mandatory and non-mandatory appendices to the articles; i.e., (1) Scope, roles and responsibilities, (2) Materials, (3) Structural design, (4) Fabrication and installation, (5) Non-destructive examination, (6) Pressure and leak testing, and (7) Terms used in general requirements. (author)

Source term evaluation for accident transients in the experimental fusion facility ITER

Energy Technology Data Exchange (ETDEWEB)

Virot, F.; Barrachin, M.; Cousin, F. [IRSN, BP3-13115, Saint Paul lez Durance (France)

2015-03-15

We have studied the transport and chemical speciation of radio-toxic and toxic species for an event of water ingress in the vacuum vessel of experimental fusion facility ITER with the ASTEC code. In particular our evaluation takes into account an assessed thermodynamic data for the beryllium gaseous species. This study shows that deposited beryllium dusts of atomic Be and Be(OH){sub 2} are formed. It also shows that Be(OT){sub 2} could exist in some conditions in the drain tank. (authors)

Fusion neutronics plan in the development of fusion reactor. With the aim of realizing electric power

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hiroo; Morimoto, Yuichi; Ochiai, Kentarou; Sugimoto, Masayoshi; Nishitani, Takeo; Takeuchi, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-10-01

On June 1992, Atomic Energy Commission in Japan has settled Third Phase Program of Fusion Research and Development to achieve self-ignition condition, to realize long pulse burning plasma and to establish basis of fusion engineering for demonstration reactor. This report describes research plan of Fusion Neutron Laboratory in JAERI toward a development of fusion reactor with an aim of realizing electric power. The fusion neutron laboratory has a fusion neutronics facility (FNS), intense fusion neutron source. The plan includes research items in the FNS; characteristics of shielding and breeding materials, nuclear characteristics of materials, fundamental irradiation process of insulator, diagnostics materials and structural materials, and development of in-vessel diagnostic technology. Upgrade of the FNS is also described. Also, the International Fusion Material Irradiation Facility (IFMIF) for intense neutron source to develop fusion materials is described. (author)

Accelerator conceptual design of the international fusion materials irradiation facility

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, M.; Kinsho, M. [Japan Atomic Energy Res. Inst., Tokai, Ibaraki (Japan). Intense Neutron Source Lab.; Jameson, R.A.; Blind, B. [Los Alamos National Lab., NM (United States); Teplyakov, V. [Institute for High Energy Physics, Moscow (Russian Federation); Berwald, D.; Bruhwiler, D.; Peakock, M.; Rathke, J. [Northrop Grumman Corp., Bethpage, NY (United States); Deitinghoff, H.; Klein, H.; Pozimski, Y.; Volk, K. [Johann Wolfgang Goethe Univ., Frankfurt (Germany). Inst. fur Angewandte Phys.; Ferdinand, R.; Lagniel, J.-M. [CEA Saclay LNS, Gif-sur-Yvette (France); Miyahara, A. [Teikyo Univ., Tokyo (Japan); Olivier, M. [CEA DSM, Saclay, Gif-sur-Yvette (France); Piechowiak, E. [Northrop Grumman Corp., Baltimore, MD (United States); Tanabe, Y. [Toshiba Corp., Tsurumi-ku, Yokohama (Japan)

1998-10-01

The accelerator system of the international fusion materials irradiation facility (IFMIF) provides the 250-mA, 40-MeV continuous-wave deuteron beam at one of the two lithium target stations. It consists of two identical linear accelerator modules, each of which independently delivers a 125-mA beam to the common footprint of 20 cm x 5 cm at the target surface. The accelerator module consists of an ion injector, a 175 MHz RFQ and eight DTL tanks, and rf power supply system. The requirements for the accelerator system and the design concept are described. The interface issues and operational considerations to attain the proposed availability are also discussed. (orig.) 8 refs.

The National Ignition Facility and the Promise of Inertial Fusion Energy

International Nuclear Information System (INIS)

Moses, E.I.

2010-01-01

The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in Livermore, CA, is now operational. The NIF is the world's most energetic laser system capable of producing 1.8 MJ and 500 TW of ultraviolet light. By concentrating the energy from its 192 extremely energetic laser beams into a mm 3 -sized target, NIF can produce temperatures above 100 million K, densities of 1,000 g/cm 3 , and pressures 100 billion times atmospheric pressure - conditions that have never been created in a laboratory and emulate those in planetary interiors and stellar environments. On September 29, 2010, the first integrated ignition experiment was conducted, demonstrating the successful coordination of the laser, cryogenic target system, array of diagnostics and infrastructure required for ignition demonstration. In light of this strong progress, the U.S. and international communities are examining the implication of NIF ignition for inertial fusion energy (IFE). A laser-based IFE power plant will require a repetition rate of 10-20 Hz and a laser with 10% electrical-optical efficiency, as well as further development and advances in large-scale target fabrication, target injection, and other supporting technologies. These capabilities could lead to a prototype IFE demonstration plant in the 10- to 15-year time frame. LLNL, in partnership with other institutions, is developing a Laser Inertial Fusion Engine (LIFE) concept and examining in detail various technology choices, as well as the advantages of both pure fusion and fusion-fission schemes. This paper will describe the unprecedented experimental capabilities of the NIF and the results achieved so far on the path toward ignition. The paper will conclude with a discussion about the need to build on the progress on NIF to develop an implementable and effective plan to achieve the promise of LIFE as a source of carbon-free energy.

The National Ignition Facility and the Promise of Inertial Fusion Energy

Energy Technology Data Exchange (ETDEWEB)

Moses, E I

2010-12-13

The National Ignition Facility (NIF) at the Lawrence Livermore National Laboratory (LLNL) in Livermore, CA, is now operational. The NIF is the world's most energetic laser system capable of producing 1.8 MJ and 500 TW of ultraviolet light. By concentrating the energy from its 192 extremely energetic laser beams into a mm{sup 3}-sized target, NIF can produce temperatures above 100 million K, densities of 1,000 g/cm{sup 3}, and pressures 100 billion times atmospheric pressure - conditions that have never been created in a laboratory and emulate those in planetary interiors and stellar environments. On September 29, 2010, the first integrated ignition experiment was conducted, demonstrating the successful coordination of the laser, cryogenic target system, array of diagnostics and infrastructure required for ignition demonstration. In light of this strong progress, the U.S. and international communities are examining the implication of NIF ignition for inertial fusion energy (IFE). A laser-based IFE power plant will require a repetition rate of 10-20 Hz and a laser with 10% electrical-optical efficiency, as well as further development and advances in large-scale target fabrication, target injection, and other supporting technologies. These capabilities could lead to a prototype IFE demonstration plant in the 10- to 15-year time frame. LLNL, in partnership with other institutions, is developing a Laser Inertial Fusion Engine (LIFE) concept and examining in detail various technology choices, as well as the advantages of both pure fusion and fusion-fission schemes. This paper will describe the unprecedented experimental capabilities of the NIF and the results achieved so far on the path toward ignition. The paper will conclude with a discussion about the need to build on the progress on NIF to develop an implementable and effective plan to achieve the promise of LIFE as a source of carbon-free energy.

Relativistic self focussing of laser beams at fast ignitor inertial fusion with volume ignition

International Nuclear Information System (INIS)

Osman, F.; Castillo, R.; Hora, H.

1999-01-01

The alternative to the magnetic confinement fusion is inertial fusion energy mostly using lasers as drivers for compression and heating of pellets with deuterium and tritium fuel. Following the present technology of lasers with pulses of some megajoules energy and nanosecond duration, a power station for very low cost energy production (and without the problems of well erosion of magnetic confinement) could be available within 15 to 20 years. For the pellet compression, the scheme of spark ignition was mostly applied but its numerous problems with asymmetries and instabilities may be overcome by the alternative scheme of high gain volume ignition. This is a well established option of inertial fusion energy with lasers where a large range of possible later improvements is implied with respect to laser technology or higher plasma compression leading to energy production of perhaps five times below the present lowest level cost from fission reactors. A further improvement may be possible by the recent development of lasers with picosecond pulse duration using the fast igniter scheme which may reach even higher fusion gains with laser pulse energies of some 100 kilojoules

Maintenance and waste treatment of tritium existing in and out of the fusion reactor systems. 6. Study of tritium confinement in the facility of a fusion reactor

International Nuclear Information System (INIS)

Kobayashi, Kazuhiro

2000-01-01

In a future fusion reactor, tritium confinement is one of the key issues for safety. Large amount of tritium (a few grams to a hundred grams level) has been handled safely at the several facilities in the world for fusion research under the multiple confinement concept. In this chapter, the study of tritium behavior in large space such as the building is described using the Caisson Assembly for Tritium Safety (CATS) study such as the final confinement and the present R and D status concerning the tritium confinement is reviewed. (author)

Electrical energy and cost for the mirror fusion test facility

International Nuclear Information System (INIS)

Pence, G.

1983-01-01

An operational scenario has been developed for the Mirror Fusion Test Facility (MFTF-B) based on the System Requirements, our experience with existing systems, and discussions with the project engineers and designers who are responsible for the systems. This scenario was used to predict the amount of electrical energy needed for running the facility. A generic type listing is included for the equipment considered in each system. A figure shows the anticipated power drain during a five-minute shot sequence from the 115-kV substation, and from the 230-kV and direct feed substations. At this time, the three major substations that will be used for the MFTF-B are billed under three different rate schedules. A table lists these schedules and what they are anticipated as being when the facility becomes operational. The system availability, which is expected to be 0.7 or better, has not been factored into these calculations. This gives a worst case cost for the MFTF-B. Based on this study, it appears that our energy bill will be over $500 000 per month, on the average. This expenditure will constitute a significant portion of the budget needed to operate the MFTF-B. As the systems are refined, and a more accurate picture is obtained as to the size and operational cycles of the equipment, this report will be updated

Progress of nuclear fusion research and review on development of fusion reactors

International Nuclear Information System (INIS)

1976-01-01

Set up in October 1971, the ad hoc Committee on Survey of Nuclear Fusion Reactors has worked on overall fusion reactor aspects and definition of the future problems under four working groups of core, nuclear heat, materials and system. The presect volume is intended to provide reference materials in the field of fusion reactor engineering, prepared by members of the committee. Contents are broadly the following: concept of the nuclear fusion reactor, fusion core engineering, fusion reactor blanket engineering, fusion reactor materials engineering, and system problems in development of fusion reactors. (Mori, K.)

Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

International Nuclear Information System (INIS)

Sugimoto, Masayoshi

2001-01-01

International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

Seismic analysis of the mirror fusion test facility building

International Nuclear Information System (INIS)

Coats, D.W.

1978-01-01

This report describes a seismic analysis of the present Mirror Fusion Test Facility (MFTF) building at the Lawrence Livermore Laboratory. The analysis was conducted to evaluate how the structure would withstand the postulated design-basis earthquake (DBE). We discuss the methods of analysis used and results obtained. Also presented are a detailed description of the building, brief discussions of site geology, seismicity, and soil conditions, the approach used to postulate the DBE, and two methods for incorporating the effects of ductility. Floor spectra for the 2nd, 3rd, and 4th floors developed for preliminary equipment design are also included. The results of the analysis, based on best-estimate equipment loadings, indicate additional bracing and upgrading of connection details are required for the structure to survive the postulated design-basis earthquake. Specific recommendations are made

Association Between Treatment at High-Volume Facilities and Improved Overall Survival in Soft Tissue Sarcomas.

Science.gov (United States)

Venigalla, Sriram; Nead, Kevin T; Sebro, Ronnie; Guttmann, David M; Sharma, Sonam; Simone, Charles B; Levin, William P; Wilson, Robert J; Weber, Kristy L; Shabason, Jacob E

2018-03-15

Soft tissue sarcomas (STS) are rare malignancies that require complex multidisciplinary management. Therefore, facilities with high sarcoma case volume may demonstrate superior outcomes. We hypothesized that STS treatment at high-volume (HV) facilities would be associated with improved overall survival (OS). Patients aged ≥18 years with nonmetastatic STS treated with surgery and radiation therapy at a single facility from 2004 through 2013 were identified from the National Cancer Database. Facilities were dichotomized into HV and low-volume (LV) cohorts based on total case volume over the study period. OS was assessed using multivariable Cox regression with propensity score-matching. Patterns of care were assessed using multivariable logistic regression analysis. Of 9025 total patients, 1578 (17%) and 7447 (83%) were treated at HV and LV facilities, respectively. On multivariable analysis, high educational attainment, larger tumor size, higher grade, and negative surgical margins were statistically significantly associated with treatment at HV facilities; conversely, black race and non-metropolitan residence were negative predictors of treatment at HV facilities. On propensity score-matched multivariable analysis, treatment at HV facilities versus LV facilities was associated with improved OS (hazard ratio, 0.87, 95% confidence interval, 0.80-0.95; P = .001). Older age, lack of insurance, greater comorbidity, larger tumor size, higher tumor grade, and positive surgical margins were associated with statistically significantly worse OS. In this observational cohort study using the National Cancer Database, receipt of surgery and radiation therapy at HV facilities was associated with improved OS in patients with STS. Potential sociodemographic disparities limit access to care at HV facilities for certain populations. Our findings highlight the importance of receipt of care at HV facilities for patients with STS and warrant further study into improving access to

Mechanical behavior of the mirror fusion test Facility superconducting magnet coils

International Nuclear Information System (INIS)

Horvath, J.A.

1980-01-01

The mechanical response to winding and electromagnetic loads of the Mirror Fusion Test Facility (MFTF) superconducting coil pack is presented. The 375-ton (3300 N) MFTF Yin-Yang magnet, presently the world's largest superconducting magnet, is scheduled for acceptance cold-testing in May of 1981. The assembly is made up of two identical coils which together contain over 15 miles (24 km) of superconductor wound in 58 consecutive layers of 24 turns each. Topics associated with mechanical behavior include physical properties of the coil pack and its components, winding pre-load effects, finite element analysis, magnetic load redistribution, and the design impact of predicted conductor motion

The National Ignition Facility. The path to ignition and inertial fusion energy

International Nuclear Information System (INIS)

Eric Storm

2010-01-01

Complete text of publication follows. The National Ignition Facility (NIF), the world's largest and most energetic laser system built for studying inertial confinement fusion (ICF) and high-energy-density (HED) science, is now operational at Lawrence Livermore National Laboratory (LLNL). NIF's 192 beams are capable of producing 1.8 MJ and 500 TW of ultraviolet light and are configured to create pressures as high as 100 GB, matter temperatures approaching 10 9 and densities over 1000 g/cm 3 . With these capabis70lities, the NIF will enable exploring scientific problems in strategic defense, basic science and fusion energy. One of the early NIF campaigns is focusing on demonstrating laboratory-scale thermonuclear ignition and burn to produce net fusion energy gains of 10-20 with 1.2 to 1.4 MJ of 0.35 μm light. NIF ignition experiments began late in FY2009 as part of the National Ignition Campaign (NIC). Participants of NIC include LLNL, General Atomics, Los Alamos National Laboratory, Sandia National Laboratory, and the University of Rochester Laboratory for Energetics (LLE) as well as variety of national and international collaborators. The results from these initial experiments show great promise for the relatively near-term achievement of ignition. Capsule implosion experiments at energies up to 1.2 MJ have demonstrated laser energetics, radiation temperatures, and symmetry control that scale to ignition conditions. Of particular importance is the demonstration of peak hohlraum temperatures near 300 eV with low overall backscatter less than 10%. Cryogenic target capability and additional diagnostics are being installed in preparation for layered target deuterium-tritium implosions to be conducted later in 2010. The goal for NIC is to demonstrate a predictable fusion experimental platform by the end of 2012. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of Inertial Fusion Energy (IFE) and

Proposed rf system for the fusion materials irradiation test facility

International Nuclear Information System (INIS)

Fazio, M.V.; Johnson, H.P.; Hoffert, W.J.; Boyd, T.J.

1979-01-01

Preliminary rf system design for the accelerator portion of the Fusion Materials Irradiation Test (FMIT) Facility is in progress. The 35-MeV, 100-mA, cw deuteron beam will require 6.3 MW rf power at 80 MHz. Initial testing indicates the EIMAC 8973 tetrode is the most suitable final amplifier tube for each of a series of 15 amplifier chains operating at 0.5-MW output. To satisfy the beam dynamics requirements for particle acceleration and to minimize beam spill, each amplifier output must be controlled to +-1 0 in phase and the field amplitude in the tanks must be held within a 1% tolerance. These tolerances put stringent demands on the rf phase and amplitude control system

Assessment of fusion facility dose rate map using mesh adaptivity enhancements of hybrid Monte Carlo/deterministic techniques

International Nuclear Information System (INIS)

Ibrahim, Ahmad M.; Wilson, Paul P.; Sawan, Mohamed E.; Mosher, Scott W.; Peplow, Douglas E.; Grove, Robert E.

2014-01-01

Highlights: •Calculate the prompt dose rate everywhere throughout the entire fusion energy facility. •Utilize FW-CADIS to accurately perform difficult neutronics calculations for fusion energy systems. •Develop three mesh adaptivity algorithms to enhance FW-CADIS efficiency in fusion-neutronics calculations. -- Abstract: Three mesh adaptivity algorithms were developed to facilitate and expedite the use of the CADIS and FW-CADIS hybrid Monte Carlo/deterministic techniques in accurate full-scale neutronics simulations of fusion energy systems with immense sizes and complicated geometries. First, a macromaterial approach enhances the fidelity of the deterministic models without changing the mesh. Second, a deterministic mesh refinement algorithm generates meshes that capture as much geometric detail as possible without exceeding a specified maximum number of mesh elements. Finally, a weight window coarsening algorithm decouples the weight window mesh and energy bins from the mesh and energy group structure of the deterministic calculations in order to remove the memory constraint of the weight window map from the deterministic mesh resolution. The three algorithms were used to enhance an FW-CADIS calculation of the prompt dose rate throughout the ITER experimental facility and resulted in a 23.3% increase in the number of mesh tally elements in which the dose rates were calculated in a 10-day Monte Carlo calculation. Additionally, because of the significant increase in the efficiency of FW-CADIS simulations, the three algorithms enabled this difficult calculation to be accurately solved on a regular computer cluster, eliminating the need for a world-class super computer

Historical Perspective on the United States Fusion Program

International Nuclear Information System (INIS)

Dean, Stephen O.

2005-01-01

Progress and Policy is traced over the approximately 55 year history of the U. S. Fusion Program. The classified beginnings of the effort in the 1950s ended with declassification in 1958. The effort struggled during the 1960s, but ended on a positive note with the emergence of the tokamak and the promise of laser fusion. The decade of the 1970s was the 'Golden Age' of fusion, with large budget increases and the construction of many new facilities, including the Tokamak Fusion Test Reactor (TFTR) and the Shiva laser. The decade ended on a high note with the passage of the Magnetic Fusion Energy Engineering Act of 1980, overwhelming approved by Congress and signed by President Carter. The Act called for a '$20 billion, 20 year' effort aimed at construction of a fusion Demonstration Power Plant around the end of the century. The U. S. Magnetic Fusion Energy program has been on a downhill slide since 1980, both in terms of budgets and the construction of new facilities. The Inertial Confinement Fusion program, funded by Department of Energy Defense Programs, has faired considerably better, with the construction of many new facilities, including the National Ignition Facility (NIF)

A Fusion Nuclear Science Facility for a fast-track path to DEMO

Energy Technology Data Exchange (ETDEWEB)

Garofalo, A.M., E-mail: garofalo@fusion.gat.com [General Atomics, San Diego, CA (United States); Abdou, M.A. [University of California, Los Angeles, Los Angeles, CA (United States); Canik, J.M. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Chan, V.S.; Hyatt, A.W. [General Atomics, San Diego, CA (United States); Hill, D.N. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Morley, N.B. [University of California, Los Angeles, Los Angeles, CA (United States); Navratil, G.A. [Columbia University, New York, NY (United States); Sawan, M.E. [University of Wisconsin Madison, Madison, WI (United States); Taylor, T.S.; Wong, C.P.C.; Wu, W. [General Atomics, San Diego, CA (United States); Ying, A. [University of California, Los Angeles, Los Angeles, CA (United States)

2014-10-15

Highlights: • A FNSF is needed to reduce the knowledge gaps to a fusion DEMO and accelerate progress toward fusion energy. • FNSF will test and qualify first-wall/blanket components and materials in a DEMO-relevant fusion environment. • The Advanced Tokamak approach enables reduced size and risks, and is on a direct path to an attractive target power plant. • Near term research focus on specific tasks can enable starting FNSF construction within the next ten years. - Abstract: An accelerated fusion energy development program, a “fast-track” approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 m, κ = 2.3, B{sub T} = 5.4 T, I{sub P} = 6.6 MA, β{sub N} = 2.75, P{sub fus} = 127 MW. The modest bootstrap fraction of ƒ{sub BS} = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ∼ 10 in ITER.

Present status of the EPFL (Swiss) fusion-fission experiment 'LOTUS'

International Nuclear Information System (INIS)

Haldy, P.A.; Frueh, R.; Ligou, J.; Schneeberger, J.P.; Kumar, A.

1984-01-01

The present status of the LOTUS project - a fusion-fission hybrid research facility under construction at the Ecole Polytechnique Federale de Lausanne (EPFL) Switzerland - is presented. Emphasis is places on the description of the facility and on the design studies of an initial blanket of the ''fission-suppressed'' type. The LOTUS facility consists of a parallelepiped-shaped blanket, occupying roughly a volume of 1 m 3 , driven by a sealed 14 MeV (D,T) neutron generator with a rated source strength of 5x10 12 n/s. The experiment is housed in a massive concrete shielding of 220 cm thick walls, which leaves an experimental test chamber of 360 cm by 240 cm lateral dimensions and a height of 300 cm. (orig.) [de

Engineering design of a fusion test reactor (FTR) and fusion engineering research facility (FERF) based on a toroidal theta pinch

International Nuclear Information System (INIS)

Abdou, M.; Burke, R.J.; Dauzvardis, P.V.; Foss, M.; Gerstl, S.A.W.; Maroni, V.A.; Pierce, A.W.; Turner, A.F.; Krakowski, R.A.; Linford, R.K.; Oliphant, T.A.; Ribe, F.L.; Thomassen, K.I.

1975-01-01

This paper describes two advanced toroidal theta-pinch devices which are being proposed for future construction. The Fusion Test Reactor (FTR) is being designed to produce thermonuclear energy (at 20 MeV/neutron) equal to the maximum plasma energy (Q=1) and to demonstrate α-particle heating. The Fusion Engineering and Research Facility (FERF) is being designed to test materials in a fusion environment where the average 14-MeV neutron flux from the plasma is greater than or of the order of 5.10 13 n/cm 2 .s over large surface areas. These devices employ the staged theta-pinch principle where the heating is accomplished by rapid (about 0.1 μs) implosion and expansion followed by a slow compression of the plasma. The rapid implosion injects as much heat as possible at as large a plasma radious as possible so that the plasma remains stable even after further compression. The final compression to ignition requires the transfer of a large amount of magnetic energy which implies a long transfer time (about 1 ms) for realistic voltages in the driving circuit. Throughout the heating and burn cycle the plasma must remain in equilibrium and stable to the dominant MHD-modes. A sufficiently large plasma radius guarantees stability against the m = 1 modes. These equilibrium and stability conditions and the requirements on thermonuclear burn determine the design parameters for either machine. The design parameters must also be consistent with economic limitations and technological feasibility of components. In addition to these requirements, the FERF must provide a steady and reliable source of fusion neutrons. (author)

Making of the NSTX Facility

International Nuclear Information System (INIS)

Neumeyer, C.; Ono, M.; Kaye, S.M.; Peng, Y.-K.M.

1999-01-01

The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, the first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations

A Plan for the Development of Fusion Energy. Final Report to Fusion Energy Sciences Advisory Committee, Fusion Development Path Panel

Energy Technology Data Exchange (ETDEWEB)

None, None

2003-03-05

This report presents a plan for the deployment of a fusion demonstration power plant within 35 years, leading to commercial application of fusion energy by mid-century. The plan is derived from the necessary features of a demonstration fusion power plant and from the time scale defined by President Bush. It identifies critical milestones, key decision points, needed major facilities and required budgets.

Fusion research at Imperial College

International Nuclear Information System (INIS)

Haines, M.G.

1990-01-01

The historical roots of fusion research at Imperial College can be traced back to 1946 with the pioneering work of G.P. Thomson. At present research in fusion is carried out in several research groups with interdisciplinary work managed by the Centre for Fusion Studies. The principal research activity will be centred on a newly funded 5 TW pulsed power facility allowing an experimental and theoretical study of radiation collapse and fusion conditions in the dense Z-pinch. Laser-plasma studies relevant to inertial confinement are carried out using the Rutherford-Appleton Laboratory's Central Laser Facility and the new ultra-short pulse (300 fs) laser facility at Imperial College. There is a significant collaboration on the Joint European Torus and the Next European Torus together with a continuation of a long association with Culham Laboratory. Several European collaborations funded by the Comission of the European Communities and other world-wide collaborations form an integral part of this university programme, which is by far the largest in the UK. After a sketch of the historical development of fusion activities, the current and future programme of fusion research at Imperial College is presented in each of the three broad areas: the Z-pinch, laser-driven inertial confinement fusion and tokamak and other conventional magnetic confinement schemes. A summary of the funding and collaborations is outlined. (author)

Staged deployment of the International Fusion Materials Irradiation Facility

International Nuclear Information System (INIS)

Takeuchi, H.; Sugimoto, M.; Nakamura, H.

2001-01-01

The International Fusion Materials Irradiation Facility (IFMIF) employs an accelerator based D-Li intense neutron source as defined in the 1995-96 Conceptual Design Activity (CDA) study. In 1999, IEA mandated a review of the CDA IFMIF design for cost reduction without change to its original mission. This objective was accomplished by eliminating the previously assumed possibility of potential upgrade of IFMIF beyond the user requirements. The total estimated cost was reduced from $797.2 M to $487.8 M. An option of deployment in 3 stages was also examined to reduce the initial investment and annual expenditures during construction. In this scenario, full performance is achieved gradually with each interim stage as follows. 1st Stage: 20% operation for material selection for ITER breeding blanket, 2nd Stage: 50% operation to demonstrate materials performance of a reference alloy for DEMO, 3rd Stage: full performance operation ( 2MW/m 2 at 500cm 3 ) to obtain engineering data for potential DEMO materials under irradiation up to 100-200 dpa. In summary, the new, reduced cost IFMIF design and staged deployment still satisfies the original mission. The estimated cost of the 1st Stage facility is only $303.6 M making it financially much more attractive. Currently, IFMIF Key Element Technology Phase (KEP) is underway to reduce the key technology risk factors. (author)

Analysis of Consequences in the Loss-of-Coolant Accident in Wendelstein 7-X Experimental Nuclear Fusion Facility

Energy Technology Data Exchange (ETDEWEB)

Uspuras, E., E-mail: algis@mail.lei.lt [Laboratory of Nuclear Installations Safety, Lithuanian Energy Institute, Kaunas (Lithuania)

2012-09-15

Full text: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Starting 2007, Lithuanian energy institute (LEI) is a member of European Fusion Development Agreement (EFDA) organization. LEI is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) in the frames of EFDA project by performing safety analysis of fusion device W7-X. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. The W7-X facility divertor cooling system consists of two coolant circuits: the main cooling circuit and the so-called 'baking' circuit. Before plasma operation, the divertor and other invessel components must be heated up in order to 'clean' the surfaces by thermal desorption and the subsequent pumping out of the released volatile molecules. The rupture of pipe, providing water for the divertor targets during the 'baking' regime is one of the critical failure events, since primary and secondary steam production leads to a rapid increase of the inner pressure in the plasma (vacuum) vessel. Such initiating event could lead to the loss of vacuum condition up to overpressure of the plasma vessel, damage of in-vessel components and bellows of the ports. In this paper the safety analysis of 40 mm inner diameter coolant pipe rupture in cooling circuit and discharge of steam-water mixture through the leak into plasma vessel during the W7-X no-plasma 'baking' operation mode is presented. For the analysis the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers) and plasma vessel was developed by employing system thermal-hydraulic state-of-the-art RELAP5 Mod 3.3 code. This paper demonstrated, that the developed RELAP5 model allows to analyze the processes in divertor cooling system and plasma vessel

Vent rate of superconducting magnets during quench in the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Slack, D.S.

1979-01-01

When a superconducting magnet goes normal, resistive heating in the conductor evaporates surrounding LHe, which must be vented. The nature and speed at which the magnet goes normal and He is vented are not subject to rigorous analysis. This paper presents vent data from an existing magnet. An approximate mathematical model is derived and fitted to the data to permit scaling of vent requirements to larger size magnets. The worst case models of the vent employed in Mirror Fusion Test Facility (MFTF) cryogenic system design are also presented

Developing maintainability for tokamak fusion power systems. Phase II report. Volume I: executive summary

International Nuclear Information System (INIS)

Fuller, G.M.; Zahn, H.S.; Mantz, H.C.; Kaletta, G.R.; Waganer, L.M.; Carosella, L.A.; Conlee, J.L.

1978-11-01

The purpose of this report is to identify design features of fusion power reactors which contribute to the achievement of high levels of maintainability. Volume 1, the Executive Summary, presents the progress achieved toward this objective in this phase and includes a comparison with the results of the first phase study efforts. A series of maintainability design guidelines and an improved maintenance system are defined as initial steps in developing the requirements for a maintainable tokamak fusion power system. The principle comparative studies that are summarized include the determination of the benefits of various vacuum wall arrangements, the effect of unscheduled and scheduled maintenance of the first wall/blanket, some initial investigation of maintenance required for subsystems other than the first wall/blanket, and the impact of maintenance equipment failures

Design of a Gamma Reaction History Diagnostic for the National Ignition Facility

International Nuclear Information System (INIS)

Malone, R.M.; Cox, B.C.; Frogget, B.C.; Kaufman, M.I.; Tunnell, T.W.; Herrmann, H.W.; Evans, S.C.; Mack, J.M; Young, C.S.; Stoeffl, W.

2009-01-01

Gas Cherenkov detectors have been used to convert fusion gammas into photons to achieve gamma reaction history (GRH) measurements. These gas detectors include a converter, pressurized gas volume, relay optics, and a photon detector. A novel design for the National Ignition Facility (NIF) using 90 o Off-Axis Parabolic mirrors efficiently collects signal from fusion gammas with 8-ps time dispersion.1 Fusion gammas are converted to Compton electrons, which generate broadband Cherenkov light (our response is from 250 to 700 nm) in a pressurized gas cell. This light is relayed into a high-speed detector using three parabolic mirrors. The detector optics collect light from a 125-mm-diameter by 600-mm-long interchangeable gas (CO2 or SF6) volume. Because light is collected from source locations throughout the gas volume, the detector is positioned at the stop position rather than at an image position. The stop diameter and its position are independent of the light-generation locations along the gas cell. This design incorporates a fixed time delay that allows the detector to recover from prompt radiation. Optical ray tracings demonstrate how light can be collected from different angled trajectories of the Compton electrons as they traverse the gas volume. A Monte Carlo model of the conversion process from gammas to Cherenkov photons is used to generate photon trajectories. The collection efficiencies for different gamma energies are evaluated. At NIF, a cluster of four channels will allow for increased dynamic range, as well as different gamma energy thresholds. This GRH design is compared to a gas Cherenkov detector that utilizes a Cassegrain reflector now used at the OMEGA laser facility. 1. R. M. Malone, H. W. Herrmann, W. Stoeffl, J. M. Mack, C. S. Young, 'Gamma bang time/reaction history diagnostics for the National Ignition Facility using 90 o off-axis parabolic mirrors', Rev. Sci. Instrum. 79, 10E532 (2008)

Antares facility for inertial-fusion experiments: status and plans

International Nuclear Information System (INIS)

Goldstone, P.D.; Allen, G.; Jansen, H.; Saxman, A.; Singer, S.; Thuot, M.

1982-01-01

Antares is a large, 30 to 40 kJ CO 2 laser system which will provide a base for experiments to determine the efficiency with which 10 μm light can be used to drive target implosions while maintaining an acceptable level of preheat. Construction of the facility is in the final stages and diagnostics for initial experiments are being designed and constructed with operations scheduled to begin early in FY-84. After an initial shakedown period, we expect to perform a series of measurements to determine the energy scaling of hot electron temperature and target coupling efficiency in selected set of targets including simple spheres. We also expect to continue experiments, now planned for Helios, to determine whether CO 2 -produced ions are appropriate for driving inertial fusion targets with acceptable efficiency (Helios experiments have demonstrated that as much as 40% of the incident light can be converted to fast ions). Details of these experiments, as well as plans for further experiments, are still being defined

LLE 1998 annual report, October 1997 -September 1998. Inertial fusion program and National Laser Users' Facility program

International Nuclear Information System (INIS)

1999-01-01

This report summarizes research at the Laboratory for Laser Energetics (LLE), the operation of the National Laser Users' Facility (NLUF), and programs involving the education of high school, undergraduate, and graduate students for FY98. Research summaries cover: progress in laser fusion; diagnostic development; laser and optical technology; and advanced technology for laser targets

LLE 1998 annual report, October 1997--September 1998. Inertial fusion program and National Laser Users` Facility program

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-01-01

This report summarizes research at the Laboratory for Laser Energetics (LLE), the operation of the National Laser Users` Facility (NLUF), and programs involving the education of high school, undergraduate, and graduate students for FY98. Research summaries cover: progress in laser fusion; diagnostic development; laser and optical technology; and advanced technology for laser targets.

Development scenario for laser fusion

International Nuclear Information System (INIS)

Maniscalco, J.A.; Hovingh, J.; Buntzen, R.R.

1976-01-01

This scenario proposes establishment of test and engineering facilities to (1) investigate the technological problems associated with laser fusion, (2) demonstrate fissile fuel production, and (3) demonstrate competitive electrical power production. Such facilities would be major milestones along the road to a laser-fusion power economy. The relevant engineering and economic aspects of each of these research and development facilities are discussed. Pellet design and gain predictions corresponding to the most promising laser systems are presented for each plant. The results show that laser fusion has the potential to make a significant contribution to our energy needs. Beginning in the early 1990's, this new technology could be used to produce fissile fuel, and after the turn of the century it could be used to generate electrical power

World Energy Data System (WENDS). Volume VII. Nuclear facility profiles, AG--CH. [Brief tabulated information

Energy Technology Data Exchange (ETDEWEB)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile.

World Energy Data System (WENDS). Volume VIII. Nuclear facility profiles, CO--HU. [Brief tabulated information

Energy Technology Data Exchange (ETDEWEB)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile.

World Energy Data System (WENDS). Volume IX. Nuclear facility profiles, IN--PL. [Brief tabulated information

Energy Technology Data Exchange (ETDEWEB)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile.

World Energy Data System (WENDS). Volume X. Nuclear facility profiles, PO--ZA. [Brief tabulated information

Energy Technology Data Exchange (ETDEWEB)

1979-06-01

In this compendium each profile of a nuclear facility is a capsule summary of pertinent facts regarding that particular installation. The facilities described include the entire fuel cycle in the broadest sense, encompassing resource recovery through waste management. Power plants and all US facilities have been excluded. To facilitate comparison the profiles have been recorded in a standard format. Because of the breadth of the undertaking some data fields do not apply to the establishment under discussion and accordingly are blank. The set of nuclear facility profiles occupies four volumes; the profiles are ordered by country name, and then by facility code. Each nuclear facility profile volume contains two complete indexes to the information. The first index aggregates the facilities alphabetically by country. It is further organized by category of facility, and then by the four-character facility code. It provides a quick summary of the nuclear energy capability or interest in each country and also an identifier, the facility code, which can be used to access the information contained in the profile.

Gettering high energy plasma in the end loss region of the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Goldner, A.I.; Margolies, D.S.

1979-01-01

The ions escaping from the end loss fan of the Mirror Fusion Test Facility (MFTF) neutralize when they hit the surface of the end dome. If the neutrals then bounce back into the oncoming plasma, they are likely to reionize, drawing power from the center of the plasma and reducing the overall electron temperature. In this paper we describe two methods for reducing the reionization rate and a computer code for estimating their effectiveness

ROK-PRC Cooperation on Laser Fusion Energy

International Nuclear Information System (INIS)

Rhee, Yong Joo; Han, J. M.; Lee, S. M.; Nam, S. M.; Kwan, D. H.; Cha, Y. H.; Baek, S. H.

2009-03-01

International treaties on the reduction of green-house gases are now being established worldwide and Korea is supposed to join these treaties in a near future. Meanwhile the energy production via fission reactors proposed as a solution to this global environmental contamination has still inherent problems in that it also produces long-life radioactive nuclear waste in the long run, causing many serious social issues. Now the ultimate solution in this situation is believed to be the production of energy by the nuclear fusion reaction. In this project, the collaboration regarding high energy laser fusion has been carried out mainly at the Chinese facility such as ShengGuang II (SG II) laser facility, and ultrahigh intensity laser system of KAERI has been used for the small scale laser fusion and production of fast neutrons. Thomson scattering experiment to analyze the fusion plasma, opacity measurement to understand and develop the computer simulation techniques have been carried out at SG II facility, and experiments on implosion reaction which is basic to laser fusion as well as that of X-ray absorption and transmission have been done at the GEKKO XII facility of ILE, Japan. Satisfactory results both for Korea and China have been deduced by the strategy of project such that different approaches for high energy laser fusion and low energy laser fusion were applied. That is, Korean partner could get opportunities of doing experiments at the large laser facilities to get plasma diagnostic technologies and high density simulation technologies, besides the opportunity to participate in the K-C-J collaborative experiments of implosion and X-ray spectroscopy. And Chinese partner could solve their problem related to the laser fusion and neutron generation which were not successful even with their far high 300TW laser system

Fire hazard analysis for fusion energy experiments

International Nuclear Information System (INIS)

Alvares, N.J.; Hasegawa, H.K.

1979-01-01

The 2XIIB mirror fusion facility at Lawrence Livermore Laboratory (LLL) was used to evaluate the fire safety of state-of-the-art fusion energy experiments. The primary objective of this evaluation was to ensure the parallel development of fire safety and fusion energy technology. Through fault-tree analysis, we obtained a detailed engineering description of the 2XIIB fire protection system. This information helped us establish an optimum level of fire protection for experimental fusion energy facilities as well as evaluate the level of protection provided by various systems. Concurrently, we analyzed the fire hazard inherent to the facility using techniques that relate the probability of ignition to the flame spread and heat-release potential of construction materials, electrical and thermal insulations, and dielectric fluids. A comparison of the results of both analyses revealed that the existing fire protection system should be modified to accommodate the range of fire hazards inherent to the 2XIIB facility

Quasi-elastic scattering an alternative tool for mapping the fusion barriers for heavy-ion induced fusion reaction

International Nuclear Information System (INIS)

Behera, B.R.

2016-01-01

Heavy element synthesis through heavy-ion induced fusion reaction is an active field in contemporary nuclear physics. Exact knowledge of fusion barrier is one of the essential parameters for planning any experiments for heavy element production. Theoretically there are many models available to predict the exact barrier. Though these models are successful for predicting the fusion of medium mass nuclei, it somehow fails for predicting the exact location of barrier for fusion of heavy nuclei. Experimental determination of barrier for such reactions is required for future experiments for the synthesis of heavy elements. Traditionally fusion barrier is determined taking a double derivative of fusion excitation function. However, such method is difficult in case of fusion of heavy nuclei due to its very low fusion/capture cross section and its experimental complications. Alternatively fusion barrier can be determined by measuring the quasi-elastic cross section at backward angles. This method can be applied for determining the fusion barrier for the fusion of heavy nuclei. Experimental determination of fusion barrier by different methods and comparison of the fusion excitation function and quasi-elastic scattering methods for the determination of fusion barrier are reviewed. At IUAC, New Delhi recently a program has been started for the measurement of fusion barrier through quasi-elastic scattering methods. The experimental facility and the first results of the experiments carried out with this facility are presented. (author)

Fusion Canada issue 12

International Nuclear Information System (INIS)

1990-10-01

A short bulletin from the National Fusion Program. Included in this issue is a report on Darlington's Tritium Removal Facility, work at universities on Deuterium Diffusivity in Beryllium, Fusion Studies, confinement research and the operation of divertors at Tokamak de Varennes. 5 figs

Analysis of the accident with the coolant discharge into the plasma vessel of the W7-X fusion experimental facility

Energy Technology Data Exchange (ETDEWEB)

Ušpuras, E.; Kaliatka, A.; Kaliatka, T., E-mail: tadas@mail.lei.lt

2013-06-15

Highlights: • The accident with water ingress into the plasma vessel in Wendelstein nuclear fusion device W7-X was analyzed. • The analysis of the processes in the plasma vessel and ventilation system was performed using thermal-hydraulic RELAP5 Mod3.3 code. • The suitability of pressure increase prevention system was assessed. • All analyses results will be used for the optimization of W7-X design and to ensure safe operation of this nuclear fusion device. -- Abstract: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Starting 2007, Lithuanian Energy Institute (LEI) is a member of European Fusion Development Agreement (EFDA) organization. LEI is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) in the frames of EFDA project by performing safety analysis of fusion device W7-X. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. In this paper the safety analysis of 40 mm inner diameter coolant pipe rupture in cooling circuit and discharge of steam–water mixture through the leak into plasma vessel during the W7-X no-plasma “baking” operation mode is presented. For the analysis the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers) and plasma vessel was developed by employing system thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code. This paper demonstrated that the developed RELAP5 model enables to analyze the processes in divertor cooling system and plasma vessel. The results of analysis demonstrated that the proposed burst disc, connecting the plasma vessel with venting system, opens and pressure inside plasma vessel does not exceed the limiting 1.1 × 10{sup 5} Pa absolute pressure. Thus, the plasma vessel remains intact after loss

IFMIF, International Fusion Materials Irradiation Facility conceptual design activity cost report

International Nuclear Information System (INIS)

Rennich, M.J.

1996-12-01

This report documents the cost estimate for the International Fusion Materials Irradiation Facility (IFMIF) at the completion of the Conceptual Design Activity (CDA). The estimate corresponds to the design documented in the Final IFMIF CDA Report. In order to effectively involve all the collaborating parties in the development of the estimate, a preparatory meeting was held at Oak Ridge National Laboratory in March 1996 to jointly establish guidelines to insure that the estimate was uniformly prepared while still permitting each country to use customary costing techniques. These guidelines are described in Section 4. A preliminary cost estimate was issued in July 1996 based on the results of the Second Design Integration Meeting, May 20--27, 1996 at JAERI, Tokai, Japan. This document served as the basis for the final costing and review efforts culminating in a final review during the Third IFMIF Design Integration Meeting, October 14--25, 1996, ENEA, Frascati, Italy. The present estimate is a baseline cost estimate which does not apply to a specific site. A revised cost estimate will be prepared following the assignment of both the site and all the facility responsibilities

Inertial fusion science in Europe

International Nuclear Information System (INIS)

Bigot, B.

2006-01-01

Europe has built significant laser facilities to study inertial confinement fusion since the beginning of this science. The goal is to understand the processes of ignition and propagation of thermonuclear combustion. Three routes toward fusion are pursued, each of which has advantages and difficulties. The conventional routes are using a central hot spot created by the same compression and heating laser beams, either with indirect or direct drive. A more recent route, 'fast ignition', has been actively studied since the 90's, increasing the need for very high energy lasers to create the hot spot; some European lasers of this kind are already functioning, others are under construction or planned. Among European facilities, Laser Mega Joule (LMJ), which is under construction, will be the most powerful tool at the end of the decade, along with NIF in the Usa, to study and obtain fusion. LMJ is designed not only to obtain fusion but also to carry out experiments on all laser-plasma physics themes thanks to its flexibility. This facility, mainly dedicated to defence programmes, will be accessible to the academic research community. On all these facilities, numerous results are and will be obtained in the fields of High Energy Density Physics and Ultra High Intensity. (author)

Fusion Canada issue 12

Energy Technology Data Exchange (ETDEWEB)

NONE

1990-10-01

A short bulletin from the National Fusion Program. Included in this issue is a report on Darlington`s Tritium Removal Facility, work at universities on Deuterium Diffusivity in Beryllium, Fusion Studies, confinement research and the operation of divertors at Tokamak de Varennes. 5 figs.

SUPER-FMIT, an accelerator-based neutron source for fusion components irradiation testing

International Nuclear Information System (INIS)

Burke, R.J.; Holmes, J.J.; Johnson, D.L.; Mann, F.M.; Miles, R.R.

1984-01-01

The SUPER-FMIT facility is proposed as an advanced accelerator based neutron source for high flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. There, neutrons would be produced by a 0.1 ampere beam of 35 MeV deuterons incident upon a liquid lithium target. The volume available for high flux (> 10 14 n/cm 2 -s) testing in SUPER-FMIT would be 14 liters, about a factor of 30 larger than in the FMIT facility. This is because the effective beam current of 35 MeV deuterons on target can be increased by a factor of ten to 1.0 amperes or more. Such a large increase can be accomplished by acceleration of multiple beams of molecular deuterium ions (D 2 +) to 70 MeV in a common accelerator sructure. The availability of multiple beams and large total current allows great variety in the testing that can be done. For example, fluxes greater than 10 16 n/cm 2 -s, multiple simultaneous experiments, and great flexibility in tailoring of spatial distributions of flux and spectra can be achieved

Open-ended fusion devices and reactors

International Nuclear Information System (INIS)

Kawabe, T.; Nariai, H.

1983-01-01

Conceptual design studies on fusion reactors based upon open-ended confinement schemes, such as the tandem mirror and rf plugged cusp, have been carried out in Japan. These studies may be classified into two categories: near-term devices (Fusion Engineering Test Facility), and long-term fusion power recators. In the first category, a two-component cusp neutron source was proposed. In the second category, the GAMMA-R, a tandem-mirror power reactor, and the RFC-R, an axisymetric mirror and cusp, reactor studies are being conducted at the University of Tsukuba and the Institute of Plasma Physics. Mirror Fusion Engineering Facility parameters and a schematic are shown. The GAMMA-R central-cell design schematic is also shown

IFMIF (International Fusion Materials Irradiation Facility) key element technology phase task description

Energy Technology Data Exchange (ETDEWEB)

Ida, M.; Nakamura, H.; Sugimoto, M.; Yutani, T.; Takeuchi, H. [eds.] [Japan Atomic Energy Research Inst., Tokai Research Establishment, Fusion Neutron Laboratory, Tokai, Ibaraki (Japan)

2000-08-01

In 2000, a 3 year Key Element technology Phase (KEP) of the International Fusion Materials Irradiation Facility (IFMIF) has been initiated to reduce the key technology risk factors needed to achieve continuous wave (CW) beam with the desired current and energy and to reach the corresponding power handling capabilities in the liquid lithium target system. In the KEP, the IFMIF team (EU, Japan, Russian Federation, US) will perform required tasks. The contents of the tasks are described in the task description sheet. As the KEP tasks, the IFMIF team have proposed 27 tasks for Test Facilities, 12 tasks for Target, 26 tasks for Accelerator and 18 tasks for Design Integration. The task description by RF is not yet available. The task items and task descriptions may be added or revised with the progress of KEP activities. These task description sheets have been compiled in this report. After 3 years KEP, the results of the KEP tasks will be reviewed. Following the KEP, 3 years Engineering Validation Phase (EVP) will continue for IFMIF construction. (author)

Activity measurements at a waste volume reduction facility

International Nuclear Information System (INIS)

Richardson, J.; Lee, D.A.

1979-01-01

The monitoring program for Ontario Hydro's radioactive waste management site will be described, several aspects of which will be discussed in detail. The program at this facility includes categorization, volume reduction processing, and storage of solid radioactive wastes from nuclear generating stations of the CANDU type. At the present time, two types of volume reduction process are in operation - incineration and compaction. Following categorization and processing, wastes are stored in in-ground concrete trenches or tile-holes, or in above-ground quadricells. The monitoring program is divided into three areas: public safety, worker safety, and structural integrity. Development projects with respect to the monitoring program have been undertaken to achieve activity accounting for the total waste management program. In particular, a field measurement for the radioactivity content of radioactive ash containers and compacted waste drums

Design and fabrication of the superconducting-magnet system for the Mirror Fusion Test Facility (MFTF-B)

International Nuclear Information System (INIS)

Tatro, R.E.; Wohlwend, J.W.; Kozman, T.A.

1982-01-01

The superconducting magnet system for the Mirror Fusion Test Facility (MFTF-B) consists of 24 magnets; i.e. two pairs of C-shaped Yin-Yang coils, four C-shaped transition coils, four solenoidal axicell coils, and a 12-solenoid central cell. General Dynamics Convair Division has designed all the coils and is responsible for fabricating 20 coils. The two Yin-Yang pairs (four coils) are being fabricated by the Lawrence Livermore National Laboratory. Since MFTF-B is not a magnet development program, but rather a major physics experiment critical to the mirror fusion program, the basic philosophy has been to use proven materials and analytical techniques wherever possible. The transition and axicell coils are currently being analyzed and designed, while fabrication is under way on the solenoid magnets

Fusion Engineering Device design description

International Nuclear Information System (INIS)

Flanagan, C.A.; Steiner, D.; Smith, G.E.

1981-12-01

The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein

Fusion engineering device design description

Energy Technology Data Exchange (ETDEWEB)

Flanagan, C.A.; Steiner, D.; Smith, G.E.

1981-12-01

The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein.

Fusion engineering device design description

International Nuclear Information System (INIS)

Flanagan, C.A.; Steiner, D.; Smith, G.E.

1981-12-01

The US Magnetic Fusion Engineering Act of 1980 calls for the operation of a Fusion Engineering Device (FED) by 1990. It is the intent of the Act that the FED, in combination with other testing facilities, will establish the engineering feasibility of magnetic fusion energy. During 1981, the Fusion Engineering Design Center (FEDC), under the guidance of a Technical Management Board (TMB), developed a baseline design for the FED. This design is summarized herein

Laser fusion experiments, facilities and diagnostics at Lawrence Livermore Laboratory

International Nuclear Information System (INIS)

Ahlstrom, H.G.

1980-02-01

The progress of the LLL Laser Fusion Program to achieve high gain thermonuclear micro-explosions is discussed. Many experiments have been successfully performed and diagnosed using the large complex, 10-beam, 30 TW Shiva laser system. A 400 kJ design of the 20-beam Nova laser has been completed. The construction of the first phase of this facility has begun. New diagnostic instruments are described which provide one with new and improved resolution, information on laser absorption and scattering, thermal energy flow, suprathermal electrons and their effects, and final fuel conditions. Measurements were made on the absorption and Brillouin scattering for target irradiations at both 1.064 μm and 532 nm. These measurements confirm the expected increased absorption and reduced scattering at the shorter wavelength. Implosion experiments have been performed which have produced final fuel densities over the range of 10x to 100x liquid DT density

Field-reversal experiments in the mirror fusion test facility (MFTF)

International Nuclear Information System (INIS)

Shearer, J.W.; Condit, W.C.

1977-01-01

Detailed consideration of several aspects of a field-reversal experiment was begun in the Mirror Fusion Test Facility (MFTF): Model calculations have provided some plausible parameters for a field-reversed deuterium plasma in the MFTF, and a buildup calculation indicates that the MFTF neutral-beam system is marginally sufficient to achieve field reversal by neutral injection alone. However, the many uncertainties indicate the need for further research and development on alternate buildup methods. A discussion of experimental objectives is presented and important diagnostics are listed. The range of parameter space accessible with the MFTF magnet design is explored, and we find that with proper aiming of the neutral beams, meaningful experiments can be performed to advance toward these objectives. Finally, it is pointed out that if we achieve enhanced n tau confinement by means of field reversal, then quasi-steady-state operation of MFTF is conceivable

Data triggered data processing at the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Jackson, R.J.; Balch, T.R.; Preckshot, G.G.

1986-01-01

A primary characteristic of most batch systems is that the input data files must exist before jobs are scheduled. On the Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory the authors schedule jobs to process experimental data to be collected during a five minute shot cycle. The data driven processing system emulates a coarsely granular data flow architecture. Processing jobs are scheduled before the experimental data is collected. Processing jobs ''fire'', or execute, as input data becomes available. Similar to UNIX ''pipes'', data produced by upstream processing nodes may be used as inputs by following nodes. Users, working on the networked SUN workstations, specify data processing templates which define processes and their data dependencies. Data specifications indicate the source of data; actual associations with specific data instantiations are made when the jobs are scheduled. The authors report here on details of diagnostic data processing and their experiences

Coil winder for the magnet of the mirror fusion test facility

International Nuclear Information System (INIS)

Ling, R.C.

1977-01-01

A coil winder was designed for the purpose of fabricating the superconducting magnets of the Mirror Fusion Test Facility. The superconducting magnets are a displaced ying-yang pair, each having major and minor radii of 2.5 and 0.75 m, respectively, and cross section of 0.42 m by about 1.03 m. The superconductor cross section is a square, 13 mm on a side, and consists of a core of niobium-titanium embedded copper and a solid copper stabilizer. Conceptual studies made at Lawrence Livermore Laboratory of the coil winder resulted in concept drawings and a procurement specification. Final design was made by the contractor, and the coil winder is now in fabrication. This paper describes the performance requirements of the winder, and the evolution of its design from conceptual stage to completion

Comparison of tritium production facilities

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2002-01-01

Detailed investigation and research on the source of tritium, tritium production facilities and their comparison are presented based on the basic information about tritium. The characteristics of three types of proposed tritium production facilities, i.e., fissile type, accelerator production tritium (APT) and fusion type, are presented. APT shows many advantages except its rather high cost; fusion reactors appear to offer improved safety and environmental impacts, in particular, tritium production based on the fusion-based neutron source costs much lower and directly helps the development of fusion energy source

Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

International Nuclear Information System (INIS)

Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won; Cho, Seungyon

2014-01-01

The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity

Performance Test of Korea Heat Load Test Facility (KoHLT-EB) for the Plasma Facing Components of Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Suk-Kwon; Jin, Hyung Gon; Lee, Eo Hwak; Yoon, Jae-Sung; Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The main components of the plasma facing components (PFCs) in the tokamak are the blanket first wall and divertor, which include the armour materials, the heat sink with the cooling mechanism, and the diagnostics devices for the temperature measurement. The Korea Heat Load Test facility by using electron beam (KoHLT-EB) has been operating for the plasma facing components to develop fusion engineering. This electron beam facility was constructed using a 300 kW electron gun and a cylindrical vacuum chamber. Performance tests were carried out for the calorimetric calibrations with Cu dummy mockup and for the heat load test of large Cu module. For the simulation of the heat load test of each mockup, the preliminary thermal-hydraulic analyses with ANSYS-CFX were performed. For the development of the plasma facing components in the fusion reactors, test mockups were fabricated and tested in the high heat flux test facility. To perform a beam profile test, an assessment of the possibility of electron beam Gaussian power density profile and the results of the absorbed power for that profile before the test starts are needed. To assess the possibility of a Gaussian profile, for the qualification test of the Gaussian heat load profile, a calorimeter mockup and large Cu module were manufactured to simulate real heat. For this high-heat flux test, the Korean high-heat flux test facility using an electron beam system was constructed. In this facility, a cyclic heat flux test will be performed to measure the surface heat flux, surface temperature profile, and cooling capacity.

Siting, design and cost of shallow land burial facilities in northern New England. Volume 1

International Nuclear Information System (INIS)

1985-05-01

This study investigated the technical feasibility and cost of shallow land burial (SLB) as one low-level radioactive waste disposal option for Maine and the northern New England states of Maine, New Hampshire, and Vermont. The results are presented in five chapters addressing the licensing process for an SLB facility, the siting process, the engineering design, the cost of disposal, and the cost of transportation. Chapter 2 reviews the Federal and State licensing processes and requirements for development of an SLB facility. Included in this discussion are the stages in the life cycle of SLB facility. Chapter 3 provides site selection criteria for Maine and presents a proposed site selection methodology. The site selection criteria are defined and the reasoning behind their selection is explained. Chapter 4 discusses SLB trench and facility designs and costs. To accommodate different waste volume scenarios, differently sized facilities are discussed, representing Maine going-it-alone and a northern New England compact. Designs and costs of scenarios including nuclear power plant decommissioning wastes are also discussed. Cost estimates of licensing, facility construction, operation, closure, and post closure care are presented for the different waste volume scenarios. Chapter 5 presents estimates of what it would cost LLW generators to dispose of their waste in a Maine-only or a northern New England shallow land burial facility. The reliability of the estimates and their sensitivity to changes in waste volume are also discussed. Chapter 6 examines transportation costs

Hazard evaluation of The International Fusion Materials Irradiation Facility

Energy Technology Data Exchange (ETDEWEB)

Burgazzi, Luciano [ENEA-Centro Ricerche ' Ezio Clementel' , Advanced Physics Technology Division, Via Martiri di Monte Sole, 4, 40129 Bologna (Italy)]. E-mail: burgazzi@bologna.enea.it

2005-01-15

The International Fusion Materials Irradiation Facility (IFMIF) is aimed to provide an intense neutron source by a high current deuteron linear accelerator and a high-speed lithium flow target, for testing candidate materials for fusion. Liquid lithium is being circulated through a loop and is kept at a temperature above its freezing point. In the frame of the design phase called Key Element technology Phase (KEP), jointly performed by an international team to verify the most important risk factors, safety assessment of the whole plant has been required in order to identify the hazards associated with the plant operation. This paper discusses the safety assessments that were performed and their outcome: Failure Mode and Effect Analysis (FMEA) approach has been adopted in order to accomplish the task. Main conclusions of the study is that, on account of the safety and preventive measures adopted, potential plant related hazards are confined within the IFMIF security boundaries and great care must be exercised to protect workers and site personnel from operating the plant. The analysis has provided as a result a set of Postulated Initiating Events (PIEs), that is off-normal events, that could result in hazardous consequences for the plant, together with the total frequency and the list of component failures which could induce the PIE: this assures the exhaustive list of major initiating events of accident sequences, helpful to the further accident sequence analysis phase. Finally, for each one of the individuated PIEs, the evaluation of the accident evolution, in terms of effects on the plant and relative countermeasures, has allowed to verify that adequate measures are being taken both to prevent the accident occurrence and to cope with the accident consequences, thus assuring the fulfilment of the safety requirements.

Fusion research principles

CERN Document Server

Dolan, Thomas James

2013-01-01

Fusion Research, Volume I: Principles provides a general description of the methods and problems of fusion research. The book contains three main parts: Principles, Experiments, and Technology. The Principles part describes the conditions necessary for a fusion reaction, as well as the fundamentals of plasma confinement, heating, and diagnostics. The Experiments part details about forty plasma confinement schemes and experiments. The last part explores various engineering problems associated with reactor design, vacuum and magnet systems, materials, plasma purity, fueling, blankets, neutronics

Volume navigation with contrast enhanced ultrasound and image fusion for percutaneous interventions: first results.

Directory of Open Access Journals (Sweden)

Ernst Michael Jung

Full Text Available OBJECTIVE: Assessing the feasibility and efficiency of interventions using ultrasound (US volume navigation (V Nav with real time needle tracking and image fusion with contrast enhanced (ce CT, MRI or US. METHODS: First an in vitro study on a liver phantom with CT data image fusion was performed, involving the puncture of a 10 mm lesion in a depth of 5 cm performed by 15 examiners with US guided freehand technique vs. V Nav for the purpose of time optimization. Then 23 patients underwent ultrasound-navigated biopsies or interventions using V Nav image fusion of live ultrasound with ceCT, ceMRI or CEUS, which were acquired before the intervention. A CEUS data set was acquired in all patients. Image fusion was established for CEUS and CT or CEUS and MRI using anatomical landmarks in the area of the targeted lesion. The definition of a virtual biopsy line with navigational axes targeting the lesion was achieved by the usage of sterile trocar with a magnetic sensor embedded in its distal tip employing a dedicated navigation software for real time needle tracking. RESULTS: The in vitro study showed significantly less time needed for the simulated interventions in all examiners when V Nav was used (p<0.05. In the study involving patients, in all 10 biopsies of suspect lesions of the liver a histological confirmation was achieved. We also used V Nav for a breast biopsy (intraductal carcinoma, for a biopsy of the abdominal wall (metastasis of ovarial carcinoma and for radiofrequency ablations (4 ablations. In 8 cases of inflammatory abdominal lesions 9 percutaneous drainages were successfully inserted. CONCLUSION: Percutaneous biopsies and drainages, even of small lesions involving complex access pathways, can be accomplished with a high success rate by using 3D real time image fusion together with real time needle tracking.

Validation of thermal hydraulic codes for fusion reactors safety

International Nuclear Information System (INIS)

Sardain, P.; Gulden, W.; Massaut, V.; Takase, K.; Merill, B.; Caruso, G.

2006-01-01

A significant effort has been done worldwide on the validation of thermal hydraulic codes, which can be used for the safety assessment of fusion reactors. This work is an item of an implementing agreement under the umbrella of the International Energy Agency. The European part is supported by EFDA. Several programmes related to transient analysis in water-cooled fusion reactors were run in order to assess the capabilities of the codes to treat the main physical phenomena governing the accidental sequences related to water/steam discharge into the vacuum vessel or the cryostat. The typical phenomena are namely the pressurization of a volume at low initial pressure, the critical flow, the flashing, the relief into an expansion volume, the condensation of vapor in a pressure suppression system, the formation of ice on a cryogenic structure, the heat transfer between walls and fluid in various thermodynamic conditions. · A benchmark exercise has been done involving different types of codes, from homogeneous equilibrium to six equations non-equilibrium models. Several cases were defined, each one focusing on a particular phenomenon. · The ICE (Ingress of Coolant Event) facility has been operated in Japan. It has simulated an in-vessel LOCA and the discharge of steam into a pressure suppression system. · The EVITA (European Vacuum Impingement Test Apparatus) facility has been operated in France. It has simulated ingress of coolant into the cryostat, i.e. into a volume at low initial pressure containing surfaces at cryogenic temperature. This paper gives the main lessons gained from these programs, in particular the possibilities for the improvement of the computer codes, extending their capabilities. For example, the water properties have been extended below the triple point. Ice formation models have been implemented. Work has also been done on condensation models. The remaining needs for R-and-D are also highlighted. (author)

TBM/MTM for HTS-FNSF: An Innovative Testing Strategy to Qualify/Validate Fusion Technologies for U.S. DEMO

Directory of Open Access Journals (Sweden)

Laila El-Guebaly

2016-08-01

Full Text Available The qualification and validation of nuclear technologies are daunting tasks for fusion demonstration (DEMO and power plants. This is particularly true for advanced designs that involve harsh radiation environment with 14 MeV neutrons and high-temperature operating regimes. This paper outlines the unique qualification and validation processes developed in the U.S., offering the only access to the complete fusion environment, focusing on the most prominent U.S. blanket concept (the dual cooled PbLi (DCLL along with testing new generations of structural and functional materials in dedicated test modules. The venue for such activities is the proposed Fusion Nuclear Science Facility (FNSF, which is viewed as an essential element of the U.S. fusion roadmap. A staged blanket testing strategy has been developed to test and enhance the DCLL blanket performance during each phase of FNSF D-T operation. A materials testing module (MTM is critically important to include in the FNSF as well to test a broad range of specimens of future, more advanced generations of materials in a relevant fusion environment. The most important attributes for MTM are the relevant He/dpa ratio (10–15 and the much larger specimen volumes compared to the 10–500 mL range available in the International Fusion Materials Irradiation Facility (IFMIF and European DEMO-Oriented Neutron Source (DONES.

The management of fusion waste

International Nuclear Information System (INIS)

Hancox, R.; Butterworth, G.J.

1990-01-01

Fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste as a result of neutron irradiation of the structural materials and absorption of the tritium fuel. An important issue is whether the volume of this waste and the risks associated with it can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. Information is presented on the radioactive waste expected from the decommissioning of three generations of fusion devices - the JET experiment, NET, and power reactors. The characteristics and probable volumes of this waste are considered, together with the risks associated with its disposal. (author)

The management of fusion waste

International Nuclear Information System (INIS)

Hancox, R.; Butterworth, G.J.

1991-01-01

Fusion reactors based on the deuterium-tritium fuel cycle will generate radioactive waste as a result of neutron irradiation of the structural materials and absorption of the tritium fuel. An important issue is whether the volume of this waste and the risks associated with it can be reduced to a sufficiently low level that the environmental advantage of fusion can be maintained without incurring unacceptable additional costs. Information is presented on the radioactive waste expected from the decommissioning of three generations of fusion devices - the JET experiment, NET, and power reactors. The characteristics and probable volumes of this waste are considered, together with the risks associated with its disposal. (orig.)

Nuclear fusion research in Australia

International Nuclear Information System (INIS)

Cheetham, A.D.

1997-01-01

In this paper the recently formed National Plasma Fusion Research Facility centred around the H-1NF Heliac, located at the Australian National University, the Institute of Advanced Studies is described in the context of the international Stellarator program and the national collaboration with the Australian Fusion Research Group. The objectives of the facility and the planned physics research program over the next five years are discussed and some recent results will be presented. The facility will support investigations in the following research areas: finite pressure equilibrium and stability, transport in high temperature plasmas, plasma heating and formation, instabilities and turbulence, edge plasma physics and advanced diagnostic development

Fusion tritium program in Japan

International Nuclear Information System (INIS)

Okamoto, M.; Yoshida, H.; Naruse, Y.

1988-01-01

Nuclear Fusion Council, Atomic Energy Commission of Japan, has started to review the nuclear fusion R and D plan for the next stage, post JT-60. The council launched a subcommittee on fundamental issues in the nuclear fusion development in 1985, for review of the basic strategy of a development plan. The subcommittee presented an interim report in Feb. 1986 after 6 months discussion and the report was approved by the Nuclear Fusion Council. Two major R and D programs described in the interim report are the development of a Tokamak type large facility and the comprehensive development of the fusion reactor technology. The latter means to promote the reactor technologies which will be essential in the future to construct not only a D/T burning but also a DEMO reactor. The Nuclear Fusion Development Program in Japan is shown. The interim report recommended to organize two subcommittees to establish an integrated national R and D plan; one was for the design of the next step large facility and the other was for the R and program of the fusion technology. The subcommittee for the latter consisted of 7 working groups; one of them was organized for the tritium technology

Adequacy of power-to-volume scaling philosophy to simulate natural circulation in Integral Test Facilities

International Nuclear Information System (INIS)

Nayak, A.K.; Vijayan, P.K.; Saha, D.; Venkat Raj, V.; Aritomi, Masanori

1998-01-01

Theoretical and experimental investigations were carried out to study the adequacy of power-to-volume scaling philosophy for the simulation of natural circulation and to establish the scaling philosophy applicable for the design of the Integral Test Facility (ITF-AHWR) for the Indian Advanced Heavy Water Reactor (AHWR). The results indicate that a reduction in the flow channel diameter of the scaled facility as required by the power-to-volume scaling philosophy may affect the simulation of natural circulation behaviour of the prototype plants. This is caused by the distortions due to the inability to simulate the frictional resistance of the scaled facility. Hence, it is recommended that the flow channel diameter of the scaled facility should be as close as possible to the prototype. This was verified by comparing the natural circulation behaviour of a prototype 220 MWe Indian PHWR and its scaled facility (FISBE-1) designed based on power-to-volume scaling philosophy. It is suggested from examinations using a mathematical model and a computer code that the FISBE-1 simulates the steady state and the general trend of transient natural circulation behaviour of the prototype reactor adequately. Finally the proposed scaling method was applied for the design of the ITF-AHWR. (author)

Radiological safety design considerations for fusion research experiments

International Nuclear Information System (INIS)

Crase, K.W.; Singh, M.S.

1979-01-01

A wide variety of fusion research experiments are in the planning or construction stages. Two such experiments, the Nova Laser Fusion Facility and the Mirror Fusion Test Facility (MFTF), are currently under construction at Lawrence Livermore Laboratory. Although the plasma chamber vault for MFTF and the Nova target room will have thick concrete walls and roofs, the radiation safety problems are made complex by the numerous requirements for shield wall penetrations. This paper addresses radiation safety considerations for the MFTF and Nova experiments, and the need for integrated safety considerations and safety technology development during the planning stages of fusion experiments

Analysis of neutron propagation from the skyshine port of a fusion neutron source facility

Energy Technology Data Exchange (ETDEWEB)

Wakisaka, M. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan); Kaneko, J. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan)]. E-mail: kin@qe.eng.hokudai.ac.jp; Fujita, F. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan); Ochiai, K. [Japan Atomic Energy Institute, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Nishitani, T. [Japan Atomic Energy Institute, Tokai-mura, Ibaraki-ken 319-1195 (Japan); Yoshida, S. [Tokai University, 1117 Kitakaname, Hirastuka, Kanagawa-ken 259-1292 (Japan); Sawamura, T. [Hokkaido University, Kita-8, Nishi-5, Kita-ku, Sapporo 080-8628 (Japan)

2005-12-01

The process of neutron leaking from a 14MeV neutron source facility was analyzed by calculations and experiments. The experiments were performed at the Fusion Neutron Source (FNS) facility of the Japan Atomic Energy Institute, Tokai-mura, Japan, which has a port on the roof for skyshine experiments, and a {sup 3}He counter surrounded with a polyethylene moderator of different thicknesses was used to estimate the energy spectra and dose distributions. The {sup 3}He counter with a 3-cm-thick moderator was also used for dose measurements, and the doses evaluated by the counter counts and the calculated count-to-dose conversion factor agreed with the calculations to within {approx}30%. The dose distribution was found to fit a simple analytical expression, D(r)=Q{sub D}exp(-r/{lambda}{sub D})r and the parameters Q{sub D} and {lambda}{sub D} are discussed.

Preoperational baseline and site characterization report for the Environmental Restoration Disposal Facility. Volume 2, Revision 2

International Nuclear Information System (INIS)

Weekes, D.C.; Lindsey, K.A.; Ford, B.H.; Jaeger, G.K.

1996-12-01

This document is Volume 2 in a two-volume series that comprise the site characterization report, the Preoperational Baseline and Site Characterization Report for the Environmental Restoration Disposal Facility. Volume 1 contains data interpretation and information supporting the conclusions in the main text. This document presents original data in support of Volume 1 of the report. The following types of data are presented: well construction reports; borehole logs; borehole geophysical data; well development and pump installation; survey reports; preoperational baseline chemical data and aquifer test data. Five groundwater monitoring wells, six deep characterization boreholes, and two shallow characterization boreholes were drilled at the Environmental Restoration Disposal Facility (ERDF) site to directly investigate site-specific hydrogeologic conditions

Data analysis software tools for enhanced collaboration at the DIII-D National Fusion Facility

International Nuclear Information System (INIS)

Schachter, J.; Peng, Q.; Schissel, D.P.

2000-01-01

Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is 'GAPlotObj', an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is 'MDSplus', which is used at DIII-D as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are 'ReviewPlus' and 'EFITviewer'. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server

Data Analysis Software Tools for Enhanced Collaboration at the DIII-D National Fusion Facility

International Nuclear Information System (INIS)

Schachter, J.; Peng, Q.; Schissel, D.P.

1999-01-01

Data analysis at the DIII-D National Fusion Facility is simplified by the use of two software packages in analysis codes. The first is GAP1otObj, an IDL-based object-oriented library used in visualization tools for dynamic plotting. GAPlotObj gives users the ability to manipulate graphs directly through mouse and keyboard-driven commands. The second software package is MDSplus, which is used at DIED as a central repository for analyzed data. GAPlotObj and MDSplus reduce the effort required for a collaborator to become familiar with the DIII-D analysis environment by providing uniform interfaces for data display and retrieval. Two visualization tools at DIII-D that benefit from them are ReviewPlus and EFITviewer. ReviewPlus is capable of displaying interactive 2D and 3D graphs of raw, analyzed, and simulation code data. EFITviewer is used to display results from the EFIT analysis code together with kinetic profiles and machine geometry. Both bring new possibilities for data exploration to the user, and are able to plot data from any fusion research site with an MDSplus data server

Recent Accomplishments and Future Directions in US Fusion Safety & Environmental Program

Energy Technology Data Exchange (ETDEWEB)

David A. Petti; Brad J. Merrill; Phillip Sharpe; L. C. Cadwallader; L. El-Guebaly; S. Reyes

2006-07-01

The US fusion program has long recognized that the safety and environmental (S&E) potential of fusion can be attained by prudent materials selection, judicious design choices, and integration of safety requirements into the design of the facility. To achieve this goal, S&E research is focused on understanding the behavior of the largest sources of radioactive and hazardous materials in a fusion facility, understanding how energy sources in a fusion facility could mobilize those materials, developing integrated state of the art S&E computer codes and risk tools for safety assessment, and evaluating S&E issues associated with current fusion designs. In this paper, recent accomplishments are reviewed and future directions outlined.

Development of a Plasma Streaming System for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Holdsworth, T.; Clark, R.N.; McCotter, R.E.; Rossow, T.L.; Cruz, G.E.

1979-01-01

The Plasma Streaming System (PSS) is an essential portion of the Mirror Fusion Test Facility (MFTF), scheduled for completion in October 1981. The PSS will develop a plasma density of at least 2 x 10 12 particles/cm 3 at the MFTF magnet centerline by injecting particles along the field lines. The plasma will have a midplane plasma radius as large as 40 cm with variable plasma particle energy and beam geometry. Minimum amounts of impurities will be injected, with emphasis on minimizing high Z materials. Each of the 60 PSS units will consist of a gun magnet assembly (GMA) and a power supply. Each GMA consists of a plasma streaming gun, a pulse magnet that provides variable beam shaping, and a fast reaction pulse gas valve

Fusion interfaces for tactical environments: An application of virtual reality technology

Science.gov (United States)

Haas, Michael W.

1994-01-01

The term Fusion Interface is defined as a class of interface which integrally incorporates both virtual and nonvirtual concepts and devices across the visual, auditory, and haptic sensory modalities. A fusion interface is a multisensory virtually-augmented synthetic environment. A new facility has been developed within the Human Engineering Division of the Armstrong Laboratory dedicated to exploratory development of fusion interface concepts. This new facility, the Fusion Interfaces for Tactical Environments (FITE) Facility is a specialized flight simulator enabling efficient concept development through rapid prototyping and direct experience of new fusion concepts. The FITE Facility also supports evaluation of fusion concepts by operation fighter pilots in an air combat environment. The facility is utilized by a multidisciplinary design team composed of human factors engineers, electronics engineers, computer scientists, experimental psychologists, and oeprational pilots. The FITE computational architecture is composed of twenty-five 80486-based microcomputers operating in real-time. The microcomputers generate out-the-window visuals, in-cockpit and head-mounted visuals, localized auditory presentations, haptic displays on the stick and rudder pedals, as well as executing weapons models, aerodynamic models, and threat models.

Progress in high gain inertial confinement fusion

International Nuclear Information System (INIS)

Sun Jingwen

2001-01-01

The author reviews the progress in laboratory high gain inertial confinement fusion (ICF), including ICF capsule physics, high-energy-density science, inertial fusion energy, the National Ignition Facility (NIF) and its design of ignition targets and the peta watt laser breakthrough. High power laser, particle beam, and pulsed power facilities around the world have established the new laboratory field of high-energy- density plasma physics and have furthered development of inertial fusion. New capabilities such as those provided by high-brightness peta watt lasers have enabled the study of matter feasible in conditions previously unachievable on earth. Science and technology developed in inertial fusion research have found near-term commercial use and have enabled steady progress toward the goal of fusion ignition and high gain in the laboratory, and have opened up new fields of study for the 21 st century

Critical safety function guidelines for experimental fusion facilities

International Nuclear Information System (INIS)

Cadwallader, L.C.

1989-01-01

As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented

Surrogate Plant Data Base : Volume 2. Appendix C : Facilities Planning Baseline Data

Science.gov (United States)

1983-05-01

This four volume report consists of a data base describing "surrogate" automobile and truck manufacturing plants developed as part of a methodology for evaluating capital investment requirements in new manufacturing facilities to build new fleets of ...

50 years of fusion research

Science.gov (United States)

Meade, Dale

2010-01-01

Fusion energy research began in the early 1950s as scientists worked to harness the awesome power of the atom for peaceful purposes. There was early optimism for a quick solution for fusion energy as there had been for fission. However, this was soon tempered by reality as the difficulty of producing and confining fusion fuel at temperatures of 100 million °C in the laboratory was appreciated. Fusion research has followed two main paths—inertial confinement fusion and magnetic confinement fusion. Over the past 50 years, there has been remarkable progress with both approaches, and now each has a solid technical foundation that has led to the construction of major facilities that are aimed at demonstrating fusion energy producing plasmas.

Safety training and safe operating procedures written for PBFA (Particle Beam Fusion Accelerator) II and applicable to other pulsed power facilities

Energy Technology Data Exchange (ETDEWEB)

Donovan, G.L.; Goldstein, S.A.

1986-12-01

To ensure that work in advancing pulsed power technology is performed with an acceptably low risk, pulsed power research facilities at Sandia National Laboratories must satisfy general safety guidelines established by the Department of Energy, policies and formats of the Environment, Safety, and Health (ES and H) Department, and detailed procedures formulated by the Pulsed Power Sciences Directorate. The approach to safety training and to writing safe operating procedures, and the procedures presented here are specific to the Particle Beam Fusion Accelerator II (PBFA II) Facility but are applicable as guidelines to other research and development facilities which have similar hazards.

Safety training and safe operating procedures written for PBFA [Particle Beam Fusion Accelerator] II and applicable to other pulsed power facilities

International Nuclear Information System (INIS)

Donovan, G.L.; Goldstein, S.A.

1986-12-01

To ensure that work in advancing pulsed power technology is performed with an acceptably low risk, pulsed power research facilities at Sandia National Laboratories must satisfy general safety guidelines established by the Department of Energy, policies and formats of the Environment, Safety, and Health (ES and H) Department, and detailed procedures formulated by the Pulsed Power Sciences Directorate. The approach to safety training and to writing safe operating procedures, and the procedures presented here are specific to the Particle Beam Fusion Accelerator II (PBFA II) Facility but are applicable as guidelines to other research and development facilities which have similar hazards

Development of lithium target system in engineering validation and engineering design activity of the international fusion materials irradiation facility (IFMIF/EVEDA)

International Nuclear Information System (INIS)

Wakai, Eiichi; Kondo, Hiroo; Sugimoto, Masayoshi; Ida, Mizuho; Kanemura, Takuji; Watanabe, Kazuyoshi; Fujishiro, Kouji; Edao, Yuuki; Niitsuma, Shigeto; Kimura, Haruyuki; Fukada, Satoshi; Hiromoto, Tetsushi; Shigeharu, Satoshi; Yagi, Jyuro; Furukawa, Tomohiro; Hirakawa, Yasushi; Suzuki, Akihiro; Terai, Takayuki; Horiike, Hiroshi; Hoashi, Eiji; Suzuki, Sachiko; Yamaoka, Nobuo; Serizawa, Hisashi; Kawahito, Yosuke; Tsuji, Yoshiyuki; Furuya, Kazuyuki; Takeo, Fumio

2012-01-01

Engineering validation and engineering design activity (EVEDA) for the international fusion materials irradiation facility (IFMIF) has been conducted since 2007. Research and development of the Lithium target facility is an important part of this activity. We constructed a world largest liquid Lithium test loop with a capacity of 5000 L in 2010 and successfully completed the first stage validation tests (functional tests of components and Lithium flow test (flow velocity 15 m/s at the target). In the present article, recent results of the EVEDA activity for the Lithium target facility and related technologies on liquid Lithium are reviewed. (author)

Vacuum pumping of tritium in fusion power reactors

International Nuclear Information System (INIS)

Coffin, D.O.; Walthers, C.R.

1979-01-01

Compound cryopumps of three different designs will be tested with deuterium-tritium (DT) mixtures under simulated fusion reactor conditions at the Tritium Systems Test Assembly (TSTA) now being constructed at the Los Alamos Scientific Laboratory (LASL). The first of these pumps is already in operation, and its preliminary performance is presented. The supporting vacuum facility necessary to regenerate these fusion facility cryopumps is also described. The next generation of fusion system vacuum pumps may include non-cryogenic or conventional-cryogenic hybrid systems, several of which are discussed

Critical safety function guidelines for experimental fusion facilities

International Nuclear Information System (INIS)

Cadwallader, L.C.

1989-01-01

As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented. 10 refs., 6 figs

Long-Baseline Neutrino Facility (LBNF) and Deep Underground Neutrino Experiment (DUNE): Conceptual Design Report. Volume 3: Long-Baseline Neutrino Facility for DUNE

Energy Technology Data Exchange (ETDEWEB)

Strait, James [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); McCluskey, Elaine [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); Lundin, Tracy [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); Willhite, Joshua [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); Hamernik, Thomas [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); Papadimitriou, Vaia [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); Marchionni, Alberto [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); Kim, Min Jeong [National Inst. of Nuclear Physics (INFN), Frascati (Italy). National Lab. of Frascati (INFN-LNF); Nessi, Marzio [Univ. of Geneva (Switzerland); Montanari, David [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States); Heavey, Anne [Fermi National Accelerator Laboratory (FNAL), Batavia, IL (United States)

2016-01-21

This volume of the LBNF/DUNE Conceptual Design Report covers the Long-Baseline Neutrino Facility for DUNE and describes the LBNF Project, which includes design and construction of the beamline at Fermilab, the conventional facilities at both Fermilab and SURF, and the cryostat and cryogenics infrastructure required for the DUNE far detector.

Results from the CDE phase activity on neutron dosimetry for the international fusion materials irradiation facility test cell

Energy Technology Data Exchange (ETDEWEB)

Esposito, B. E-mail: esposito@frascati.enea.it; Bertalot, L.; Maruccia, G.; Petrizzi, L.; Bignan, G.; Blandin, C.; Chauffriat, S.; Lebrun, A.; Recroix, H.; Trapp, J.P.; Kaschuck, Y

2000-11-01

The international fusion materials irradiation facility (IFMIF) project deals with the study of an accelerator-based, deuterium-lithium source, producing high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials for fusion energy reactors. IFMIF would also provide calibration and validation of data from fission reactor and other accelerator based irradiation tests. This paper describes the activity on neutron/gamma dosimetry (necessary for the characterization of the specimens' irradiation) performed in the frame of the IFMIF conceptual design evaluation (CDE) neutronics tasks. During the previous phase (conceptual design activity (CDA)) the multifoil activation method was proposed for the measurement of the neutron fluence and spectrum and a set of suitable foils was defined. The cross section variances and covariances of this set of foils have now been used for tests on the sensitivity of the IFMIF neutron spectrum determination to cross section uncertainties. The analysis has been carried out using the LSL-M2 code, which optimizes the neutron spectrum by means of a least-squares technique taking into account the variance and covariance files. In the second part of the activity, the possibility of extending to IFMIF the use of existing on-line in-core neutron/gamma monitors (to be located at several positions inside the IFMIF test cell for beam control, safety and diagnostic purposes) has been studied. A feasibility analysis of the modifications required to adapt sub-miniature fission chambers (recently developed by CEA-Cadarache) to the high flux test module of the test cell has been carried out. The verification of this application pertinence and a gross definition of the in-core detector characteristics are described. The option of using self-powered neutron detectors (SPNDs) is also discussed.

Engineering test facility

International Nuclear Information System (INIS)

Steiner, D.; Becraft, W.R.; Sager, P.H.

1981-01-01

The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This paper described the design status of the ETF

Proliferation Risks of Fusion Energy: Clandestine Production, Covert Production, and Breakout

International Nuclear Information System (INIS)

Goldston, R.J.; Glaser, A.; Ross, A.F.

2009-01-01

Nuclear proliferation risks from fusion associated with access to weapon-usable material can be divided into three main categories: (1) clandestine production of fissile material in an undeclared facility, (2) covert production of such material in a declared and safeguarded facility, and (3) use of a declared facility in a breakout scenario, in which a state begins production of fissile material without concealing the effort. In this paper we address each of these categories of risk from fusion. For each case, we find that the proliferation risk from fusion systems can be much lower than the equivalent risk from fission systems, if commercial fusion systems are designed to accommodate appropriate safeguards

Inertial confinement fusion (ICF)

International Nuclear Information System (INIS)

Nuckolls, J.

1977-01-01

The principal goal of the inertial confinement fusion program is the development of a practical fusion power plant in this century. Rapid progress has been made in the four major areas of ICF--targets, drivers, fusion experiments, and reactors. High gain targets have been designed. Laser, electron beam, and heavy ion accelerator drivers appear to be feasible. Record-breaking thermonuclear conditions have been experimentally achieved. Detailed diagnostics of laser implosions have confirmed predictions of the LASNEX computer program. Experimental facilities are being planned and constructed capable of igniting high gain fusion microexplosions in the mid 1980's. A low cost long lifetime reactor design has been developed

Radiological safety design considerations for a laser-fusion facility

International Nuclear Information System (INIS)

Singh, M.S.

1977-01-01

Detailed neutronics and photonics calculations have been performed for analyzing prompt and residual radiations and required shielding associated with the design of a laser-fusion facility with a nominal yield of 10 19 neutrons per D--T burn pulse. The standard Livermore Monte Carlo codes and nuclear data cross section libraries were used in calculations. The Bateman equation was used to calculate the accumulation and decay of radionuclide chain products. A number of activation sensitivity experiments were conducted and the results were found to be in very good agreement within 10 percent of those calculated. It has been found that neutron yields of 2 x 10 19 per day can be conducted continuously if the reactor chamber is Kevlar-epoxy or silica, the primary shield is 0.60-m of water immediately on the chamber, and the building concrete is 1.80 m thick. These precautions result in dose equivalents below the primary protection limits inside the target room after a few hours of cool-down per each 10 19 pulse, 10 percent of the primary protection limits immediately outside the target room, and 1 percent of the natural background level at the nearest site boundary

Fusion safety program Annual report, Fiscal year 1995

International Nuclear Information System (INIS)

Longhurst, G.R.; Cadwallader, L.C.; Carmack, W.J.

1995-12-01

This report summarizes the major activities of the Fusion Safety Program in FY-95. 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. Among the technical areas covered in this report are tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate database development, and safety code development and application to fusion safety issues. Most 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 the technical support for commercial fusion facility conceptual design studies. A final activity described is work to develop DOE Technical Standards for Safety of Fusion Test Facilities

Data acquisition and processing system at the NOVETTE laser-fusion facility

International Nuclear Information System (INIS)

Auerbach, J.M.; Severyn, J.R.; Kroepfl, D.J.

1982-01-01

The computer hardware and software used for acquisition and processing of data from experiments at the NOVETTE laser fusion facility are described. Nearly two hundred sensors are used to measure the performance of millimeter extent targets irradiated by multi-kilojoule laser pulses. Sensor output is recorded on CAMAC based digitizers, CCD arrays, and film. CAMAC instrument outputs are acquired and collected by a network of LSI-11 microprocessors centrally controlled by a VAX 11/780. The user controls the system through menus presented on color video displays equipped with touch panels. The control VAX collects data from all microprocessors and CCD arrays and stores them in a file for transport to a second VAX 11/780 which is used for processing and final analysis. Transfer is done through a high speed fiber-optic link. Relational data bases are used extensively in the processing and archiving of data

Balanced program plan. Volume 10. Fusion: analysis for biomedical and environmental research

International Nuclear Information System (INIS)

Hungate, F.P.

1976-06-01

Development of the Balanced Program Plan for analysis for biomedical and environmental research was initiated in the spring of 1975. The goal was a redefinition of research efforts and priorities to meet ERDA's requirements for a program of health and environmental research to support the development and commercialization of energy technologies. As part of the Balanced Program planning effort the major ERDA-supported multidisciplinary laboratories were assigned responsibility for analyzing the research needs of each of nine energy technologies and describing a research program to meet these needs. The staff of the Division of Biomedical and Environmental Research was assigned the task of defining a research program addressed to each of five biomedical and environmental research categories (characterization, measurement and monitoring; physical and chemical processes and effects; health effects; ecological effects; and integrated assessment and socioeconomic processes and effects) applicable to all energy technologies. The first drafts of these documents were available for a work-shop in June 1975 at which the DBER staff and scientists from the laboratories developed a comprehensive set of program recommendations. Pacific Northwest Laboratory was assigned responsibility for defining research needs and a recommended research program for fusion and fission technologies. This report, Volume 10, presents the input for fusion

A fusion engineering program for Canada

International Nuclear Information System (INIS)

Billington, I.J.

In 1980 the National Research Council asked DSMA ATCON Ltd., in collaboration with Ontario Hydro, the University of Toronto, and McMaster University, to evaluate concepts for a national fusion engineering program, to define a facility that could be constructed in Canada to meet the program goals, and to suggest a strategy for encouraging industrial participation. The central element of the proposed fusion engineering and development program is tritium technology, with additional emphasis on the broader field of all hydrogen isotopes and their interactions with materials. The Canadian program in the initial phase would concentrate on fusion fuel systems, materials development, equipment development, and safety and the environment. A preliminary concept for the facility required has been developed, and key organizational activities identified. The total program costs should be $1 million in the first year, rising to a steady state of $5 million from the fourth year onward. The capital cost of the research facility is estimated to be $20 million spread over three years, and its operating budget around $7 million. The program as envisioned would make use of Canada's existing tritium resources and handling experience to contribute to worldwide fusion research

Local control station for development, testing and maintenance of mirror fusion facility subsystem controls

International Nuclear Information System (INIS)

Ables, E.; Kelly, M.F.

1985-01-01

A Local Control Station (LCS) was designed and built to provide a simplified ad easily configurable means of controlling any Mirror Fusion Test Facility (MFTF-B) subsystem for the purpose of development, testing and maintenance of the subsystem. All MFTF-B Subsystems incorporate at least one Local Control Computer (LCC) that is connected to and accepts high level commands from one of the Supervisory Control and Diagnostic System (SCDS) computers. The LCS connects directly to the LCC in place of SCDS. The LCS communicates with the subsystem hardware using the same SCDS commands that the local control computer recognizes and as such requires no special configuration of the LCC

Suggestions for an updated fusion power program

International Nuclear Information System (INIS)

Clarke, J.F.

1976-02-01

This document contains suggestions for a revised CTR Program strategy which should allow us to achieve equivalent goals while operating within the above constraints. The revised program is designed around three major facilities. The first is an upgrading of the present TFTR facility which will provide a demonstration of the generation of tens of megawatts electric equivalent originally envisioned for the 1985 EPR. The second device is the TTAP which will allow the integration and optimization of the plasma physics results obtained from the next generation of plasma physics experiments. The improvement in tokamak reactor operation resulting from this optimization of fusion plasma performance will enable an EPR to be designed which will produce several hundred megawatts of electric power by 1990. This will move the fusion program much closer to its goal of commercial fusion power by the turn of the century. In addition to this function the TTAP will serve as a prototype of the 1990 EPR system, thus making more certain the successful operation of this device. The third element of this revised program is an intense radiation damage facility which will provide the radiation damage information necessary for the EPR and subsequent fusion reactor facilities. The sum total of experience gained from reacting plasma experiments on TFTR, reactor grade plasma optimization and technological prototyping on TTAP, and end of life radiation damage results from the intense neutron facility will solve all of the presently foreseen problems associated with a tokamak fusion power reactor except those associated with the external nuclear systems. These external system problems such as tritium breeding and optimal power recovery can be developed in parallel on the 1990 EPR

Supervisory control and diagnostics system for the mirror fusion test facility: overview and status 1980

International Nuclear Information System (INIS)

McGoldrick, P.R.

1981-01-01

The Mirror Fusion Test Facility (MFTF) is a complex facility requiring a highly-computerized Supervisory Control and Diagnostics System (SCDS) to monitor and provide control over ten subsystems; three of which require true process control. SCDS will provide physicists with a method of studying machine and plasma behavior by acquiring and processing up to four megabytes of plasma diagnostic information every five minutes. A high degree of availability and throughput is provided by a distributed computer system (nine 32-bit minicomputers on shared memory). Data, distributed across SCDS, is managed by a high-bandwidth Distributed Database Management System. The MFTF operators' control room consoles use color television monitors with touch sensitive screens; this is a totally new approach. The method of handling deviations to normal machine operation and how the operator should be notified and assisted in the resolution of problems has been studied and a system designed

Burn-center quality improvement: are burn outcomes dependent on admitting facilities and is there a volume-outcome "sweet-spot"?

Science.gov (United States)

Hranjec, Tjasa; Turrentine, Florence E; Stukenborg, George; Young, Jeffrey S; Sawyer, Robert G; Calland, James F

2012-05-01

Risk factors of mortality in burn patients such as inhalation injury, patient age, and percent of total body surface area (%TBSA) burned have been identified in previous publications. However, little is known about the variability of mortality outcomes between burn centers and whether the admitting facilities or facility volumes can be recognized as predictors of mortality. De-identified data from 87,665 acute burn observations obtained from the National Burn Repository between 2003 and 2007 were used to estimate a multivariable logistic regression model that could predict patient mortality with reference to the admitting burn facility/facility volume, adjusted for differences in age, inhalation injury, %TBSA burned, and an additional factor, percent full thickness burn (%FTB). As previously reported, all three covariates (%TBSA burned, inhalation injury, and age) were found to be highly statistically significant risk factors of mortality in burn patients (P value improve the multivariable model. The treatment/admitting facility was found to be an independent mortality predictor, with certain hospitals having increased odds of death and others showing a protective effect (decreased odds ratio). Hospitals with high burn volumes had the highest risk of mortality. Mortality outcomes of patients with similar risk factors (%TBSA burned, inhalation injury, age, and %FTB) are significantly affected by the treating facility and their admission volumes.

Megajoule-class single-pulse KrF laser test facility as a logical step toward inertial fusion commercialization

International Nuclear Information System (INIS)

Harris, D.B.; Pendergrass, J.H.

1985-01-01

The cost and efficiency of megajoule-class KrF laser single pulse test facilities have been examined. A baseline design is described which illuminates targets with 5 MJ with shaped 10-ns pulses. The system uses 24 main amplifiers and operates with an optics operating fluence of 4.0 J/cm 2 . This system has 9.0% efficiency and costs $200/joule. Tradeoff studies indicate that large amplifier modules and high fluences lead to the lowest laser system costs, but that only a 20% cost savings can be realized by going to amplifier modules larger than 200 kJ and/or fluences greater than 4 J/cm 2 . The role of the megajoule-class single-pulse test facility towards inertial fusion commercialization will also be discussed

Facilities projects performance measurement system

International Nuclear Information System (INIS)

Erben, J.F.

1979-01-01

The two DOE-owned facilities at Hanford, the Fuels and Materials Examination Facility (FMEF), and the Fusion Materials Irradiation Test Facility (FMIT), are described. The performance measurement systems used at these two facilities are next described

Fusion-power demonstration

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Carlson, G.A.; Neef, W.S.; Moir, R.W.; Campbell, R.B.; Botwin, R.; Clarkson, I.R.; Carpenter, T.J.

1983-01-01

As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

Fusion power demonstration

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.

1983-01-01

As a satellite to the MARS (Mirror Advanced Reactor Study) a smaller, near-term device has been scoped, called the FPD (Fusion Power Demonstration). Envisioned as the next logical step toward a power reactor, it would advance the mirror fusion program beyond MFTF-B and provide an intermediate step toward commercial fusion power. Breakeven net electric power capability would be the goal such that no net utility power would be required to sustain the operation. A phased implementation is envisioned, with a deuterium checkout first to verify the plasma systems before significant neutron activation has occurred. Major tritium-related facilities would be installed with the second phase to produce sufficient fusion power to supply the recirculating power to maintain the neutral beams, ECRH, magnets and other auxiliary equipment

LIFE: The Case for Early Commercialization of Fusion Energy

International Nuclear Information System (INIS)

Anklam, T.; Simon, A.J.; Powers, S.; Meier, W.R.

2011-01-01

This paper presents the case for early commercialization of laser inertial fusion energy (LIFE). Results taken from systems modeling of the US electrical generating enterprise quantify the benefits of fusion energy in terms of carbon emission, nuclear waste and plutonium production avoidance. Sensitivity of benefits-gained to timing of market-entry is presented. These results show the importance of achieving market entry in the 2030 time frame. Economic modeling results show that fusion energy can be competitive with other low-carbon energy sources. The paper concludes with a description of the LIFE commercialization path. It proposes constructing a demonstration facility capable of continuous fusion operations within 10 to 15 years. This facility will qualify the processes and materials needed for a commercial fusion power plant.

Nuclear data requirements for fusion reactor nucleonics

International Nuclear Information System (INIS)

Bhat, M.R.; Abdou, M.A.

1980-01-01

Nuclear data requirements for fusion reactor nucleonics are reviewed and the present status of data are assessed. The discussion is divided into broad categories dealing with data for Fusion Materials Irradiation Test Facility (FMIT), D-T Fusion Reactors, Alternate Fuel Cycles and the Evaluated Data Files that are available or would be available in the near future

Inertial fusion with hypervelocity impact

International Nuclear Information System (INIS)

Olariu, S.

1998-01-01

resulting temperature being supposed sufficient to ignite the fusion reactions between the deuterium-tritium nuclei contained by the fuel target at rest. The fusion reactions thus ignited will be propagated in the entire volume of the large-yield target. In this analysis it was required that the energy produced by fusion reactions should be equal to the energy of the incident projectile. Assuming that the incident projectile is positively charged, the estimated length of the linac that would accelerate the projectile to the threshold velocity is 13 km, if deuterium-tritium densities of 570 g cm -3 could be obtained by compression. The required accelerating length is comparable to the lengths of linacs being considered in the field of high energy physics. At pressures of 10 -11 Torr in the accelerating column the heating and the ionization produced by the interaction of the incident projectile with the background gas molecules are found to be within acceptable limits. Unlike usual accelerators where a large number of particles are accelerated at a time, in the case of impact fusion the incident projectiles would be accelerated one at a time. This opens the possibility that an accelerating wave front should be generated by successively activated switches. The scale of a research facility on inertial fusion by hypervelocity impact can be reduced to that of a laboratory by using projectiles with a very small radius, of the order of 1 μm and bellow, when charge-to-mass ratio has more favorable values, and not by compressing the target at rest. The ratio of the energy released in fusion reactions to the energy of the incident projectile will in this case be much less than 1, but we still expect to see the increase of the fusion yield per incident nucleon with the radius of the projectile, which is the basic assumption of inertial impact fusion. (author)

The controlled thermonuclear fusion

International Nuclear Information System (INIS)

Barre, Bertrand

2014-01-01

After some generalities on particle physics, and on fusion and fission reactions, the author outlines that the fission reaction is easier to obtain than the fusion reaction, evokes the fusion which takes place in stars, and outlines the difficulty to manage and control this reaction: one of its application is the H bomb. The challenge is therefore to find a way to control this reaction and make it a steady and continuous source of energy. The author then presents the most promising way: the magnetic confinement fusion. He evokes its main issues, the already performed experiments (tokamak), and gives a larger presentation of the ITER project. Then, he evokes another way, the inertial confinement fusion, and the two main experimental installations (National Ignition Facility in Livermore, and the Laser Megajoule in Bordeaux). Finally, he gives a list of benefits and drawbacks of an industrial nuclear fusion

Thin shell, high velocity inertial confinement fusion implosions on the national ignition facility.

Science.gov (United States)

Ma, T; Hurricane, O A; Callahan, D A; Barrios, M A; Casey, D T; Dewald, E L; Dittrich, T R; Döppner, T; Haan, S W; Hinkel, D E; Berzak Hopkins, L F; Le Pape, S; MacPhee, A G; Pak, A; Park, H-S; Patel, P K; Remington, B A; Robey, H F; Salmonson, J D; Springer, P T; Tommasini, R; Benedetti, L R; Bionta, R; Bond, E; Bradley, D K; Caggiano, J; Celliers, P; Cerjan, C J; Church, J A; Dixit, S; Dylla-Spears, R; Edgell, D; Edwards, M J; Field, J; Fittinghoff, D N; Frenje, J A; Gatu Johnson, M; Grim, G; Guler, N; Hatarik, R; Herrmann, H W; Hsing, W W; Izumi, N; Jones, O S; Khan, S F; Kilkenny, J D; Knauer, J; Kohut, T; Kozioziemski, B; Kritcher, A; Kyrala, G; Landen, O L; MacGowan, B J; Mackinnon, A J; Meezan, N B; Merrill, F E; Moody, J D; Nagel, S R; Nikroo, A; Parham, T; Ralph, J E; Rosen, M D; Rygg, J R; Sater, J; Sayre, D; Schneider, M B; Shaughnessy, D; Spears, B K; Town, R P J; Volegov, P L; Wan, A; Widmann, K; Wilde, C H; Yeamans, C

2015-04-10

Experiments have recently been conducted at the National Ignition Facility utilizing inertial confinement fusion capsule ablators that are 175 and 165  μm in thickness, 10% and 15% thinner, respectively, than the nominal thickness capsule used throughout the high foot and most of the National Ignition Campaign. These three-shock, high-adiabat, high-foot implosions have demonstrated good performance, with higher velocity and better symmetry control at lower laser powers and energies than their nominal thickness ablator counterparts. Little to no hydrodynamic mix into the DT hot spot has been observed despite the higher velocities and reduced depth for possible instability feedthrough. Early results have shown good repeatability, with up to 1/2 the neutron yield coming from α-particle self-heating.

Radiochemical determination of Inertial Confinement Fusion capsule compression at the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Shaughnessy, D. A., E-mail: shaughnessy2@llnl.gov; Moody, K. J.; Gharibyan, N.; Grant, P. M.; Gostic, J. M.; Torretto, P. C.; Wooddy, P. T.; Bandong, B. B.; Cerjan, C. J.; Hagmann, C. A.; Caggiano, J. A.; Yeamans, C. B.; Bernstein, L. A.; Schneider, D. H. G.; Henry, E. A.; Fortner, R. J. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94551 (United States); Despotopulos, J. D. [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, California 94551 (United States); Radiochemistry Program, University of Nevada Las Vegas, Las Vegas, Nevada 89154 (United States)

2014-06-15

We describe a radiochemical measurement of the ratio of isotope concentrations produced in a gold hohlraum surrounding an Inertial Confinement Fusion capsule at the National Ignition Facility (NIF). We relate the ratio of the concentrations of (n,γ) and (n,2n) products in the gold hohlraum matrix to the down-scatter of neutrons in the compressed fuel and, consequently, to the fuel's areal density. The observed ratio of the concentrations of {sup 198m+g}Au and {sup 196g}Au is a performance signature of ablator areal density and the fuel assembly confinement time. We identify the measurement of nuclear cross sections of astrophysical importance as a potential application of the neutrons generated at the NIF.

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA

International Nuclear Information System (INIS)

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1993-03-01

This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and accelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well other topics, has been included. This volume (Volume 4 of the series) contains 209 abstracts. An author index and a subject index are provided to facilitate use. The subject index contains the abstract numbers from previous volumes, as well as the current volume

Assessment of the gas dynamic trap mirror facility as intense neutron source for fusion material test irradiations

International Nuclear Information System (INIS)

Fischer, U.; Moeslang, A.; Ivanov, A.A.

2000-01-01

The gas dynamic trap (GDT) mirror machine has been proposed by the Budker Institute of nuclear physics, Novosibirsk, as a volumetric neutron source for fusion material test irradiations. On the basis of the GDT plasma confinement concept, 14 MeV neutrons are generated at high production rates in the two end sections of the axially symmetrical central mirror cell, serving as suitable irradiation test regions. In this paper, we present an assessment of the GDT as intense neutron source for fusion material test irradiations. This includes comparisons to irradiation conditions in fusion reactor systems (ITER, Demo) and the International Fusion Material Irradiation Facility (IFMIF), as well as a conceptual design for a helium-cooled tubular test assembly elaborated for the largest of the two test zones taking proper account of neutronics, thermal-hydraulic and mechanical aspects. This tubular test assembly incorporates ten rigs of about 200 cm length used for inserting instrumented test capsules with miniaturized specimens taking advantage of the 'small specimen test technology'. The proposed design allows individual temperatures in each of the rigs, and active heating systems inside the capsules ensures specimen temperature stability even during beam-off periods. The major concern is about the maximum achievable dpa accumulation of less than 15 dpa per full power year on the basis of the present design parameters of the GDT neutron source. A design upgrading is proposed to allow for higher neutron wall loadings in the material test regions

Sustaining neutral beam power supply system for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Eckard, R.D.; Wilson, J.H.; Van Ness, H.W.

1980-01-01

In late August 1978, a fixed price procurement contract for $25,000,000 was awarded to Aydin Energy Division, Palo Alto, California, for the design, manufacture, installation and acceptance testing of the Lawrence Livermore National Laboratory Mirror Fusion Test Facility (MFTF) Sustaining Neutral Beam Power Supply System (SNBPSS). This system of 24 power supply sets will provide the conditioned power for the 24 neutral beam source modules. Each set will provide the accel potential the arc power, the filament power, and the suppressor power for its associated neutral beam source module. The design and development of the SNBPSS has progressed through the final design phase and is now in production. Testing of the major sub-assembly power supply is proceeding at Aydin and the final acceptance testing of the first two power supplies at LLNL is expected to be completed this year

Fusion Safety Program annual report, fiscal year 1994

International Nuclear Information System (INIS)

Longhurst, G.R.; Cadwallader, L.C.; Dolan, T.J.; Herring, J.S.; McCarthy, K.A.; Merrill, B.J.; Motloch, C.G.; Petti, D.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

DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-05-01

This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist.

DOE standard: Integration of environment, safety, and health into facility disposition activities. Volume 2: Appendices

International Nuclear Information System (INIS)

1998-05-01

This volume contains the appendices that provide additional environment, safety, and health (ES and H) information to complement Volume 1 of this Standard. Appendix A provides a set of candidate DOE ES and H directives and external regulations, organized by hazard types that may be used to identify potentially applicable directives to a specific facility disposition activity. Appendix B offers examples and lessons learned that illustrate implementation of ES and H approaches discussed in Section 3 of Volume 1. Appendix C contains ISMS performance expectations to guide a project team in developing and implementing an effective ISMS and in developing specific performance criteria for use in facility disposition. Appendix D provides guidance for identifying potential Applicable or Relevant and Appropriate Requirements (ARARs) when decommissioning facilities fall under the Comprehensive Environmental Response, Compensation, Liability Act (CERCLA) process. Appendix E discusses ES and H considerations for dispositioning facilities by privatization. Appendix F is an overview of the WSS process. Appendix G provides a copy of two DOE Office of Nuclear Safety Policy and Standards memoranda that form the bases for some of the guidance discussed within the Standard. Appendix H gives information on available hazard analysis techniques and references. Appendix I provides a supplemental discussion to Sections 3.3.4, Hazard Baseline Documentation, and 3.3.6, Environmental Permits. Appendix J presents a sample readiness evaluation checklist

Federal Facility Compliance Act, Proposed Site Treatment Plan: Background Volume. Executive Summary

International Nuclear Information System (INIS)

1995-01-01

This Federal Facility Compliance Act Site Treatment Plan discusses the options of radioactive waste management for Ames Laboratory. This is the background volume which discusses: site history and mission; framework for developing site treatment plans; proposed plan organization and related activities; characterization of mixed waste and waste minimization; low level mixed waste streams and the proposed treatment approach; future generation of TRU and mixed wastes; the adequacy of mixed waste storage facilities; and a summary of the overall DOE activity in the area of disposal of mixed waste treatment residuals

Canadian Fusion Fuels Technology Project annual report 93/94

Energy Technology Data Exchange (ETDEWEB)

NONE

1994-12-31

The Canadian Fusion Fuels Technology Project exists to develop fusion technologies and apply them worldwide in today`s advanced fusion projects and to apply these technologies in fusion and tritium research facilities. CFFTP concentrates on developing capability in fusion fuel cycle systems, in tritium handling technologies and in remote handling. This is an annual report for CFFTP and as such also includes a financial report.

Canadian Fusion Fuels Technology Project annual report 93/94

International Nuclear Information System (INIS)

1994-01-01

The Canadian Fusion Fuels Technology Project exists to develop fusion technologies and apply them worldwide in today's advanced fusion projects and to apply these technologies in fusion and tritium research facilities. CFFTP concentrates on developing capability in fusion fuel cycle systems, in tritium handling technologies and in remote handling. This is an annual report for CFFTP and as such also includes a financial report

Fabrication and Installation of Radiation Shielded Spent Fuel Fusion System

International Nuclear Information System (INIS)

Park, Soon Dal; Park, Yang Soon; Kim, Jong Goo; Ha, Yeong Keong; Song, Kyu Seok

2010-02-01

Most of the generated fission gases are retained in the fuel matrix in supersaturated state, thus alter the original physicochemical properties of the fuel. And some of them are released into free volume of a fuel rod and that cause internal pressure increase of a fuel rod. Furthermore, as extending fuel burnup, the data on fission gas generation(FGG) and fission gas release(FGR) are considered very important for fuel safety investigation. Consequently, it is required to establish an experimental facility for handling of highly radioactive sample and to develop an analytical technology for measurement of retained fission gas in a spent fuel. This report describes not only on the construction of a shielded glove box which can handle highly radioactive materials but also on the modifications and instrumentations of spent fuel fusion facilities and collection apparatuses of retained fission gas

Inertial Fusion Power Plant Concept of Operations and Maintenance

Energy Technology Data Exchange (ETDEWEB)

Anklam, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Knutson, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunne, A. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kasper, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sheehan, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Lang, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Roberts, V. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mau, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-01-15

Parsons and LLNL scientists and engineers performed design and engineering work for power plant pre-conceptual designs based on the anticipated laser fusion demonstrations at the National Ignition Facility (NIF). Work included identifying concepts of operations and maintenance (O&M) and associated requirements relevant to fusion power plant systems analysis. A laser fusion power plant would incorporate a large process and power conversion facility with a laser system and fusion engine serving as the heat source, based in part on some of the systems and technologies advanced at NIF. Process operations would be similar in scope to those used in chemical, oil refinery, and nuclear waste processing facilities, while power conversion operations would be similar to those used in commercial thermal power plants. While some aspects of the tritium fuel cycle can be based on existing technologies, many aspects of a laser fusion power plant presents several important and unique O&M requirements that demand new solutions. For example, onsite recovery of tritium; unique remote material handling systems for use in areas with high radiation, radioactive materials, or high temperatures; a five-year fusion engine target chamber replacement cycle with other annual and multi-year cycles anticipated for major maintenance of other systems, structures, and components (SSC); and unique SSC for fusion target waste recycling streams. This paper describes fusion power plant O&M concepts and requirements, how O&M requirements could be met in design, and how basic organizational and planning issues can be addressed for a safe, reliable, economic, and feasible fusion power plant.

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

International Nuclear Information System (INIS)

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project's scope of work. Should more accurate numbers be required, a new analysis would be necessary

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project`s scope of work. Should more accurate numbers be required, a new analysis would be necessary.

The Mirror Fusion Test Facility cryogenic system: Performance, management approach, and present equipment status

International Nuclear Information System (INIS)

Slack, D.S.; Chronis, W.C.

1987-01-01

The cryogenic system for the Mirror Fusion Test Facility (MFTF) is a 14-kW, 4.35-K helium refrigeration system that proved to be highly successful and cost-effective. All operating objectives were met, while remaining within a few percent of initial cost and schedule plans. The management approach used in MFTF allowed decisions to be made quickly and effectively, and it helped keep costs down. Manpower levels, extent and type of industrial participation, key aspects of subcontractor specifications, and subcontractor interactions are reviewed, as well as highlights of the system tests, operation, and present equipment status. Organizations planning large, high-technology systems may benefit from this experience with the MFTF cryogenic system

The engineering test facility

International Nuclear Information System (INIS)

Steiner, D.; Becraft, W.R.; Sager, P.H.

1981-01-01

The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test-bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This paper describes the design status of the ETF. (orig.)

Red Blood Cell Transfusion Need for Elective Primary Posterior Lumbar Fusion in A High-Volume Center for Spine Surgery

Science.gov (United States)

Ristagno, Giuseppe; Beluffi, Simonetta; Tanzi, Dario; Belloli, Federica; Carmagnini, Paola; Croci, Massimo; D’Aviri, Giuseppe; Menasce, Guido; Pastore, Juan C.; Pellanda, Armando; Pollini, Alberto; Savoia, Giorgio

2018-01-01

(1) Background: This study evaluated the perioperative red blood cell (RBC) transfusion need and determined predictors for transfusion in patients undergoing elective primary lumbar posterior spine fusion in a high-volume center for spine surgery. (2) Methods: Data from all patients undergoing spine surgery between 1 January 2014 and 31 December 2016 were reviewed. Patients’ demographics and comorbidities, perioperative laboratory results, and operative time were analyzed in relation to RBC transfusion. Multivariate logistic regression analysis was performed to identify the predictors of transfusion. (3) Results: A total of 874 elective surgeries for primary spine fusion were performed over the three years. Only 54 cases (6%) required RBC transfusion. Compared to the non-transfused patients, transfused patients were mainly female (p = 0.0008), significantly older, with a higher ASA grade (p = 0.0002), and with lower pre-surgery hemoglobin (HB) level and hematocrit (p < 0.0001). In the multivariate logistic regression, a lower pre-surgery HB (OR (95% CI) 2.84 (2.11–3.82)), a higher ASA class (1.77 (1.03–3.05)) and a longer operative time (1.02 (1.01–1.02)) were independently associated with RBC transfusion. (4) Conclusions: In the instance of elective surgery for primary posterior lumbar fusion in a high-volume center for spine surgery, the need for RBC transfusion is low. Factors anticipating transfusion should be taken into consideration in the patient’s pre-surgery preparation. PMID:29385760

Red Blood Cell Transfusion Need for Elective Primary Posterior Lumbar Fusion in A High-Volume Center for Spine Surgery

Directory of Open Access Journals (Sweden)

Giuseppe Ristagno

2018-01-01

Full Text Available (1 Background: This study evaluated the perioperative red blood cell (RBC transfusion need and determined predictors for transfusion in patients undergoing elective primary lumbar posterior spine fusion in a high-volume center for spine surgery. (2 Methods: Data from all patients undergoing spine surgery between 1 January 2014 and 31 December 2016 were reviewed. Patients’ demographics and comorbidities, perioperative laboratory results, and operative time were analyzed in relation to RBC transfusion. Multivariate logistic regression analysis was performed to identify the predictors of transfusion. (3 Results: A total of 874 elective surgeries for primary spine fusion were performed over the three years. Only 54 cases (6% required RBC transfusion. Compared to the non-transfused patients, transfused patients were mainly female (p = 0.0008, significantly older, with a higher ASA grade (p = 0.0002, and with lower pre-surgery hemoglobin (HB level and hematocrit (p < 0.0001. In the multivariate logistic regression, a lower pre-surgery HB (OR (95% CI 2.84 (2.11–3.82, a higher ASA class (1.77 (1.03–3.05 and a longer operative time (1.02 (1.01–1.02 were independently associated with RBC transfusion. (4 Conclusions: In the instance of elective surgery for primary posterior lumbar fusion in a high-volume center for spine surgery, the need for RBC transfusion is low. Factors anticipating transfusion should be taken into consideration in the patient’s pre-surgery preparation.

Fusion safety program plan

International Nuclear Information System (INIS)

Crocker, J.G.; Holland, D.F.; Herring, J.S.

1980-09-01

The program plan consists of research that has been divided into 13 different areas. These areas focus on the radioactive inventories that are expected in fusion reactors, the energy sources potentially available to release a portion of these inventories, and analysis and design techniques to assess and ensure that the safety risks associated with operation of magnetic fusion facilities are acceptably low. The document presents both long-term program requirements that must be fulfilled as part of the commercialization of fusion power and a five-year plan for each of the 13 different program areas. Also presented is a general discussion of magnetic fusion reactor safety, a method for establishing priorities in the program, and specific priority ratings for each task in the five-year plan

The National Ignition Facility Project

International Nuclear Information System (INIS)

Paisner, J.A.; Campbell, E.M.; Hogan, W.J.

1994-01-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. This paper reviews the design, schedule, and costs associated with the construction project

Status of tritium technology development for magnetic-fusion energy

International Nuclear Information System (INIS)

Anderson, J.L.

1983-01-01

The development of tritium technology for the magnetic fusion energy program has progressed at a rapid rate over the past two years. The focal points for this development in the United States have been the Tritium Systems Test Assembly at Los Alamos and the FED/INTOR studies supported by the Fusion Engineering Design Center at Oak Ridge. In Canada the Canadian Fusion Fuel Technology Project has been initiated and promises to make significant contributions to the tritium technology program in the next few years. The Japanese government has now approved funding for the Tritium Processing Laboratory at the Japan Atomic Energy Research Institute's Tokai Research Establishment. Construction on this new facility is scheduled to begin in April 1983. This facility will be the center for fusion tritium technology development in Japan. The European Community is currently working on the design of the tritium facility for the Joint European Torus. There is considerable interaction between all of these programs, thus accelerating the overall development of this crucial technology

A brief review of the progress of laser inertial confinement fusion in recent years

International Nuclear Information System (INIS)

Wang Ganchang

1997-01-01

The progress of laser fusion research in the world as well as in China in recent years is reviewed. A brief analysis of the main facilities of laser fusion such as National Ignition Facility in United States Omega Facility in Rochestor University and NIKE Facility in Naval Research Laboratory of United States and the experiments done on these facilities is presented

Image fusion using MIM software via picture archiving and communication system

International Nuclear Information System (INIS)

Gu Zhaoxiang; Jiang Maosong

2001-01-01

The preliminary studies of the multimodality image registration and fusion were performed using an image fusion software and a picture archiving and communication system (PACS) to explore the methodology. Original image voluminal data were acquired with a CT scanner, MR and dual-head coincidence SPECT, respectively. The data sets from all imaging devices were queried, retrieved, transferred and accessed via DICOM PACS. The image fusion was performed at the SPECT ICON work-station, where the MIM (Medical Image Merge) fusion software was installed. The images were created by re-slicing original volume on the fly. The image volumes were aligned by translation and rotation of these view ports with respect to the original volume orientation. The transparency factor and contrast were adjusted in order that both volumes can be visualized in the merged images. The image volume data of CT, MR and nuclear medicine were transferred, accessed and loaded via PACS successfully. The perfect fused images of chest CT/ 18 F-FDG and brain MR/SPECT were obtained. These results showed that image fusion technique using PACS was feasible and practical. Further experimentation and larger validation studies were needed to explore the full potential of the clinical use

Chemical processing of liquid lithium fusion reactor blankets

International Nuclear Information System (INIS)

Weston, J.R.; Calaway, W.F.; Yonco, R.M.; Hines, J.B.; Maroni, V.A.

1979-01-01

A 50-gallon-capacity lithium loop constructed mostly from 304L stainless steel has been operated for over 6000 hours at temperatures in the range from 360 to 480 0 C. This facility, the Lithium Processing Test Loop (LPTL), is being used to develop processing and monitoring technology for liquid lithium fusion reactor blankets. Results of tests of a molten-salt extraction method for removing impurities from liquid lithium have yielded remarkably good distribution coefficients for several of the more common nonmetallic elements found in lithium systems. In particular, the equilibrium volumetric distribution coefficients, D/sub v/ (concentration per unit volume of impurity in salt/concentration per unit volume of impurity in lithium), for hydrogen, deuterium, nitrogen and carbon are approx. 3, approx. 4, > 10, approx. 2, respectively. Other studies conducted with a smaller loop system, the Lithium Mini-Test Loop (LMTL), have shown that zirconium getter-trapping can be effectively used to remove selected impurities from flowing lithium

Fusion of biological membranes

Indian Academy of Sciences (India)

Home; Journals; Pramana – Journal of Physics; Volume 64; Issue 6. Fusion of biological membranes. K Katsov M Müller M Schick. Invited Talks:- Topic 11. Biologically motivated problems (protein-folding models, dynamics at the scale of the cell; biological networks, evolution models, etc.) Volume 64 Issue 6 June 2005 pp ...

The Innovations, Technology and Waste Management Approaches to Safely Package and Transport the World's First Radioactive Fusion Research Reactor for Burial

International Nuclear Information System (INIS)

Keith Rule; Erik Perry; Jim Chrzanowski; Mike Viola; Ron Strykowsky

2003-01-01

Original estimates stated that the amount of radioactive waste that will be generated during the dismantling of the Tokamak Fusion Test Reactor will approach two million kilograms with an associated volume of 2,500 cubic meters. The materials were activated by 14 MeV neutrons and were highly contaminated with tritium, which present unique challenges to maintain integrity during packaging and transportation. In addition, the majority of this material is stainless steel and copper structural metal that were specifically designed and manufactured for this one-of-a-kind fusion research reactor. This provided further complexity in planning and managing the waste. We will discuss the engineering concepts, innovative practices, and technologies that were utilized to size reduce, stabilize, and package the many unique and complex components of this reactor. This waste was packaged and shipped in many different configurations and methods according to the transportation regulations and disposal facility requirements. For this particular project, we were able to utilize two separate disposal facilities for burial. This paper will conclude with a complete summary of the actual results of the waste management costs, volumes, and best practices that were developed from this groundbreaking and successful project

Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

International Nuclear Information System (INIS)

Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

1983-01-01

The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps

Fusion Materials Irradiation Test (FMIT) facility lithium system: a design and development status

Energy Technology Data Exchange (ETDEWEB)

Brackenbury, P.J.; Bazinet, G.D.; Miller, W.C.

1983-01-01

The design and development of the Fusion Materials Irradiation Test (FMIT) Facility lithium system is outlined. This unique liquid lithium recirculating system, the largest of its kind in the world, is described with emphasis on the liquid lithium target assembly and other important components necessary to provide lithium flow to the target. The operational status and role of the Experimental Lithium System (ELS) in the design of the FMIT lithium system are discussed. Safety aspects of operating the FMIT lithium system in a highly radioactive condition are described. Potential spillage of the lithium is controlled by cell liners, by argon flood systems and by remote maintenance features. Lithium chemistry is monitored and controlled by a side-stream loop, where impurities measured by instruments are collected by hot and cold traps.

Magneto-inertial Fusion: An Emerging Concept for Inertial Fusion and Dense Plasmas in Ultrahigh Magnetic Fields

Energy Technology Data Exchange (ETDEWEB)

Thio, Francis Y.C.

2008-01-01

An overview of the U.S. program in magneto-inertial fusion (MIF) is given in terms of its technical rationale, scientific goals, vision, research plans, needs, and the research facilities currently available in support of the program. Magneto-inertial fusion is an emerging concept for inertial fusion and a pathway to the study of dense plasmas in ultrahigh magnetic fields (magnetic fields in excess of 500 T). The presence of magnetic field in an inertial fusion target suppresses cross-field thermal transport and potentially could enable more attractive inertial fusion energy systems. A vigorous program in magnetized high energy density laboratory plasmas (HED-LP) addressing the scientific basis of magneto-inertial fusion has been initiated by the Office of Fusion Energy Sciences of the U.S. Department of Energy involving a number of universities, government laboratories and private institutions.

A lower cost development path for heavy ion fusion

International Nuclear Information System (INIS)

Hogan, W.J.; Meier, W.R.

1993-01-01

If two features of the inertial fusion process are exploited successfully, they can lead to significantly lower costs for demonstrating the feasibility of commercial electric power production from this source of energy. First, fusion capsule ignition and burn physics is independent of reaction chamber size and hydrodynamically-equivalent capsules can be designed to perform at small yield, exactly as they do at large yield. This means that an integrated test of all power plant components and feasibility tests of various reaction chamber concepts can be done at much smaller sizes (about 1--2 m first wall radius) and much lower powers (tens of MWs) than magnetic fusion development facilities such as ITER. Second, the driver, which is the most expensive component of currently conceived IFE development facilities, can be used to support more than one experiment target chamber/reactor (simultaneously and/or sequentially). These two factors lead to lower development facility costs, modular facilities, and the planning flexibility to spread costs over time or do several things in parallel and thus shorten the total time needed for development of Inertial Fusion Energy (IFE). In this paper the authors describe the general feature of a heavy ion fusion development plan that takes advantage of upgradable accelerators and the ability to test chambers and reactor systems at small scale in order to reduce development time and costs

Experimental results on advanced inertial fusion schemes obtained within the HiPER project

International Nuclear Information System (INIS)

Batani, Dimitri; Santos, Jorge J.; Schurtz, Guy; Hulin, Sebastien; Ribeyre, Xavier; Nicolai, Philippe; Vauzour, Benjamin; Dorchies, Fabien; Gizzi, Leonida A.; Koester, Petra; Labate, Luca; Honrubia, Javier; Antonelli, Luca; Morace, Alessio; Volpe, Luca; Nazarov, Wiger; Pasley, John; Richetta, Maria; Lancaster, Kate; Spindloe, Christopher; Tolley, Martin; Neely, David; Kozlova, Michaela; Nejdl, Jaroslav; Rus, Bedrich; Wolowski, Jerzy; Badziak, Jan

2012-01-01

This paper presents the results of experiments conducted within the Work Package 10 (fusion experimental programme) of the HiPER project. The aim of these experiments was to study the physics relevant for advanced ignition schemes for inertial confinement fusion, i.e. the fast ignition and the shock ignition. Such schemes allow to achieve a higher fusion gain compared to the indirect drive approach adopted in the National Ignition Facility in United States, which is important for the future inertial fusion energy reactors and for realising the inertial fusion with smaller facilities. (authors)

Japanese program of materials research for fusion reactors

International Nuclear Information System (INIS)

Hasiguti, R.R.

1982-01-01

The Japanese program of materials research for fusion reactors is described based on the report to the Nuclear Fusion Council, the project research program of the Ministry of Education, Science and Culture, and other official documents. The alloy development for the first wall and its radiation damage are the main topics discussed in this paper. Materials viewpoints for the Japanese Tokamak facilities and the problems of irradiation facilities are also discussed. (orig.)

Using computer graphics to analyze the placement of neutral-beam injectors for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Horvath, J.A.

1977-01-01

To optimize the neutral-beam current incident on the fusion plasma and limit the heat load on exposed surfaces of the Mirror Fusion Test Facility magnet coils, impingement of the neutral beams on the magnet structure must be minimized. Also, placement of the neutral-beam injectors must comply with specifications for neutral-current heating of the plasma and should allow maximum flexibility to accommodate alternative beam aiming patterns without significant hardware replacement or experiment down-time. Injector placements and aimings are analyzed by means of the Structural Analysis Movie Post Processor (SAMPP), a general-purpose graphics code for the display of three-dimensional finite-element models. SAMPP is used to visually assemble, disassemble, or cut away sections of the complex three-dimensional apparatus, which is represented by an assemblage of 8-node solid finite elements. The resulting picture is used to detect and quantify interactions between the structure and the neutral-particle beams

Inertial Confinement Fusion Quarterly Report: April--June 1993. Volume 3, Number 3

Energy Technology Data Exchange (ETDEWEB)

MacGowan, B.J.; Kotowski, M.; Schleich, D. [eds.

1993-11-01

This issue of the ICF Quarterly contains six articles describing recent advances in Lawrence Livermore National Laboratory`s inertial confinement fusion (ICF) program. The current emphasis of the ICF program is in support of DOE`s National Ignition Facility (NIF) initiative for demonstrating ignition and gain with a 1-2 MJ glass laser. The articles describe recent Nova experiments and investigations tailored towards enhancing understanding of the key physics and technological issues for the NIF. Titles of the articles are: development of large-aperture KDP crystals; inner-shell photo-ionized X-ray lasers; X-ray radiographic measurements of radiation-driven shock and interface motion in solid density materials; the role of nodule defects in laser-induced damage of multilayer optical coatings; techniques for Mbar to near-Gbar equation-of-state measurements with the Nova laser; parametric instabilities and laser-beam smoothing.

ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

Science.gov (United States)

Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

2014-10-01

The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

Membrane Trafficking and Vesicle Fusion

Indian Academy of Sciences (India)

Home; Journals; Resonance – Journal of Science Education; Volume 19; Issue 5. Membrane Trafficking and Vesicle Fusion: Post-Palade Era Researchers Win the Nobel Prize. Riddhi Atul Jani Subba Rao Gangi Setty. General Article Volume 19 Issue 5 May 2014 pp 421-445 ...

Design and construction of a Gamma reaction history diagnostic for the National Ignition Facility

International Nuclear Information System (INIS)

Malone, R M; Cox, B C; Frogget, B C; Kaufman, M I; Tibbitts, A; Tunnell, T W; Evans, S C; Herrmann, H W; Kim, Y H; Mack, J M; Young, C S; McGillivray, K D; Palagi, M; Stoeffl, W

2010-01-01

Gas Cherenkov detectors have been used to convert fusion gammas into photons to record gamma reaction history measurements. These gas detectors include a converter, pressurized gas volume, relay collection optics, and a photon detector. A novel design for the National Ignition Facility (NIF) using 90 0 off-axis parabolic mirrors efficiently collects signal from fusion gammas with 8-ps time dispersion. Fusion gammas are converted to Compton electrons, which generate broadband Cherenkov light (response is from 250 to 700 nm) in a pressurized gas cell. This light is relayed into a high-speed detector using three parabolic mirrors. The relay optics collect light from a 125-mm-diameter by 600-mm-long interchangeable gas (CO 2 or SF 6 ) volume. The parabolic mirrors were electroformed instead of diamond turned to reduce scattering of the UV light. All mirrors are bare aluminum coated for maximum reflectivity. This design incorporates a 4.2-ns time delay that allows the detector to recover from prompt radiation before it records the gamma signal. At NIF, a cluster of four channels will allow for increased dynamic range, as well as different gamma energy thresholds.

World progress toward fusion energy

International Nuclear Information System (INIS)

Clarke, J.F.

1989-09-01

This paper will describe the progress in fusion science and technology from a world perspective. The paper will cover the current technical status, including the understanding of fusion's economic, environmental, and safety characteristics. Fusion experiments are approaching the energy breakeven condition. An energy gain (Q) of 30 percent has been achieved in magnetic confinement experiments. In addition, temperatures required for an ignited plasma (Ti = 32 KeV) and energy confinements about 75 percent of that required for ignition have been achieved in separate experiments. Two major facilities have started the experimental campaign to extend these results and achieve or exceed Q = 1 plasma conditions by 1990. Inertial confinement fusion experiments are also approaching thermonuclear conditions and have achieved a compression factor 100-200 times liquid D-T. Because of this progress, the emphasis in fusion research is turning toward questions of engineering feasibility. Leaders of the major fusion R and D programs in the European Community (EC), Japan, the United States, and the U.S.S.R. have agreed on the major steps that are needed to reach the point at which a practical fusion system can be designed. The United States is preparing for an experiment to address the last unexplored scientific issue, the physics of an ignited plasma, during the late 1990's. The EC, Japan, U.S.S.R., and the United States have joined together under the auspices of the International Atomic Energy Agency (IAEA) to jointly design and prepare the validating R and D for an international facility, the International Thermonuclear Experimental Reactor (ITER), to address all the remaining scientific issues and to explore the engineering technology of fusion around the turn of the century. In addition, a network of international agreements have been concluded between these major parties and a number of smaller fusion programs, to cooperate on resolving a complete spectrum of fusion science and

Tritium-management requirements for D-T fusion reactors (ETF, INTOR, FED)

International Nuclear Information System (INIS)

Finn, P.A.; Clemmer, R.G.; Misra, B.

1981-10-01

The successful operation of D-T fusion reactors will depend on the development of safe and reliable tritium-containment and fuel-recycle systems. The tritium handling requirements for D-T reactors were analyzed. The reactor facility was then designed from the viewpoint of tritium management. Recovery scenarios after a tritium release were generated to show the relative importance of various scenarios. A fusion-reactor tritium facility was designed which would be appropriate for all types of plants from the Engineering Test Facility (ETF), the International Tokamak Reactor (INTOR), and the Fusion Engineering Device (FED) to the full-scale power plant epitomized by the STARFIRE design

Magnetic-fusion energy and computers

International Nuclear Information System (INIS)

Killeen, J.

1982-01-01

The application of computers to magnetic fusion energy research is essential. In the last several years the use of computers in the numerical modeling of fusion systems has increased substantially. There are several categories of computer models used to study the physics of magnetically confined plasmas. A comparable number of types of models for engineering studies are also in use. To meet the needs of the fusion program, the National Magnetic Fusion Energy Computer Center has been established at the Lawrence Livermore National Laboratory. A large central computing facility is linked to smaller computer centers at each of the major MFE laboratories by a communication network. In addition to providing cost effective computing services, the NMFECC environment stimulates collaboration and the sharing of computer codes among the various fusion research groups

Inertial-confinement fusion with lasers

International Nuclear Information System (INIS)

Betti, R.; Hurricane, O. A.

2016-01-01

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 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 the 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

Magnetic fusion energy and computers

International Nuclear Information System (INIS)

Killeen, J.

1982-01-01

The application of computers to magnetic fusion energy research is essential. In the last several years the use of computers in the numerical modeling of fusion systems has increased substantially. There are several categories of computer models used to study the physics of magnetically confined plasmas. A comparable number of types of models for engineering studies are also in use. To meet the needs of the fusion program, the National Magnetic Fusion Energy Computer Center has been established at the Lawrence Livermore National Laboratory. A large central computing facility is linked to smaller computer centers at each of the major MFE laboratories by a communication network. In addition to providing cost effective computing services, the NMFECC environment stimulates collaboration and the sharing of computer codes among the various fusion research groups

Controlled fusion and plasma physics

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

Controlled fusion and plasma physics

International Nuclear Information System (INIS)

1995-01-01

This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC)

Controlled fusion and plasma physics

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-31

This document presents the several speeches that took place during the 22nd European Physical Society conference on Controlled Fusion and Plasma Physics in Bournemouth, UK, between the 2nd and 7th July 1995. The talks deal with new experiments carried out on several tokamaks, particularly Tore Supra, concerning plasma confinement and fusion. Some information on specific fusion devices or tokamak devices is provided, as well as results of experiments concerning plasma instability. Separate abstracts were prepared for all the 31 papers in this volume. (TEC).

Security on the US Fusion Grid

Energy Technology Data Exchange (ETDEWEB)

Burruss, Justin R.; Fredian, Tom W.; Thompson, Mary R.

2005-06-01

The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER.

Security on the US Fusion Grid

International Nuclear Information System (INIS)

Burruss, Justin R.; Fredian, Tom W.; Thompson, Mary R.

2005-01-01

The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER

Security on the US fusion grid

International Nuclear Information System (INIS)

Burruss, J.R.; Fredian, T.W.; Thompson, M.R.

2006-01-01

The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This has led to the development of the U.S. fusion grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large U.S. fusion research facilities and with users both in the U.S. and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER

The problems associated with the monitoring of complex workplace radiation fields at European high-energy accelerators and thermonuclear fusion facilities

Czech Academy of Sciences Publication Activity Database

Bilski, P.; Blomgren, J.; d´Errico, F.; Esposito, A.; Fehrenbacher, G.; Fernández, F.; Fuchs, A.; Golnik, N.; Lacoste, V.; Leuschner, A.; Sandri, S.; Silari, M.; Spurný, František; Wiegel, B.; Wright, P.

2007-01-01

Roč. 126, 1-4 (2007), s. 491-496 ISSN 0144-8420 R&D Projects: GA MŠk 1P05OC032 Grant - others:ES(XE) Contract no FI6R-012684 Institutional research plan: CEZ:AV0Z10480505 Keywords : radiation fields * european high-energy accelerators * thermonuclear fusion facilities Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 0.528, year: 2007

Assessment of fusion operators for medical imaging: application to MR images fusion

International Nuclear Information System (INIS)

Barra, V.; Boire, J.Y.

2000-01-01

We propose in the article to assess the results provided by several fusion operators in the case of T 1 - and T 2 -weighted magnetic resonance images fusion of the brain. This assessment deals with an expert visual inspection of the results and with a numerical analysis of some comparison measures found in the literature. The aim of this assessment is to find the 'best' operator according to the clinical study. This method is here applied to the quantification of brain tissue volumes on a brain phantom, and allows to select a fusion operator in any clinical study where several information is available. (authors)

Engineering test facility design center

International Nuclear Information System (INIS)

Anon.

1980-01-01

The vehicle by which the fusion program would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The ETF would provide a test bed for reactor components in the fusion environment. In order to initiate preliminary planning for the ETF decision, the Office of Fusion Energy established the ETF Design Center activity to prepare the design of the ETF. This section describes the status of this design

Fusion reactor materials

Energy Technology Data Exchange (ETDEWEB)

none,

1989-01-01

This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics.

Fusion reactor materials

International Nuclear Information System (INIS)

1989-01-01

This paper discuses the following topics on fusion reactor materials: irradiation, facilities, test matrices, and experimental methods; dosimetry, damage parameters, and activation calculations; materials engineering and design requirements; fundamental mechanical behavior; radiation effects; development of structural alloys; solid breeding materials; and ceramics

Use of the SPIRAL 2 facility for material irradiations with 14 MeV energy neutrons

International Nuclear Information System (INIS)

Mosnier, A.; Ridikas, D.; Ledoux, X.; Pellemoine, F.; Anne, R.; Huguet, Y.; Lipa, M.; Magaud, P.; Marbach, G.; Saint-Laurent, M.G.; Villari, A.C.C.

2005-01-01

The primary goal of an irradiation facility for fusion applications will be to generate a material irradiation database for the design, construction, licensing and safe operation of a fusion demonstration power station (e.g., DEMO). This will be achieved through testing and qualifying material performance under neutron irradiation that simulates service up to the full lifetime anticipated in the power plant. Preliminary investigations of 14 MeV neutron effects on different kinds of fusion material could be assessed by the SPIRAL 2 Project at GANIL (Caen, France), aiming at rare isotope beams production for nuclear physics research with first beams expected by 2009. In SPIRAL 2, a deuteron beam of 5 mA and 40 MeV interacts with a rotating carbon disk producing high-energy neutrons (in the range between 1 and 40 MeV) via C (d, xn) reactions. Then, the facility could be used for 3-4 months y -1 for material irradiation purposes. This would correspond to damage rates in the order of 1-2 dpa y -1 (in Fe) in a volume of ∼10 cm 3 . Therefore, the use of miniaturized specimens will be essential in order to effectively utilize the available irradiation volume in SPIRAL 2. Sample package irradiation temperature would be in the range of 250-1000 deg. C. The irradiation level of 1-2 dpa y -1 with 14 MeV neutrons (average energy) may be interesting for micro-structural and metallurgical investigations (e.g., mini-traction, small punch tests, etc.) and possibly for the understanding of specimen size/geometric effects of critical material properties. Due to the small test cell volume, sample in situ experiments are not foreseen. However, sample packages would be, if required, available each month after transfer in a special hot cell on-site

Summary on inertial confinement fusion

International Nuclear Information System (INIS)

Meyer-Ter-Vehn, J.

1995-01-01

Highlights on inertial confinement during the fifteenth international conference on plasma physics and controlled nuclear fusion are briefly summarized. Specifically the following topics are discussed: the US National Ignition Facility presently planned by the US Department of Energy; demonstration of diagnostics for hot spot formation; declassification of Hohlraum target design; fusion targets, in particular, the Hohlraum target design for the National Ignition Facility (NIF), Hohlraum experiments, direct drive implosions, ablative Rayleigh-Taylor instabilities, laser imprinting (of perturbations by the laser on the laser target surface), hot spot formation and mixing, hot spot implosion experiments at Lawrence Livermore National Laboratory, Livermore, USA, time resolving hot spot dynamics at the Institute of Laser Engineering (ILE), Osaka, Japan, laser-plasma interaction

Integrated operations plan for the MFTF-B Mirror Fusion Test Facility. Volume I. Organization plan

International Nuclear Information System (INIS)

1981-12-01

This plan and the accompanying MFTF-B Integrated Operations Plan are submitted in response to UC/LLNL Purchase Order 3883801, dated July 1981. The organization plan also addresses the specific tasks and trade studies directed by the scope of work. The Integrated Operations Plan, which includes a reliability, quality assurance, and safety plan and an integrated logistics plan, comprises the burden of the report. In the first section of this volume, certain underlying assumptions and observations are discussed setting the requirements and limits for organization. Section B presents the recommended structure itself. Section C Device Availability vs Maintenance and Support Efforts and Section D Staffing Levels and Skills provide backup detail and justification. Section E is a trade study on maintenance and support by LLNL staff vs subcontract and Section F is a plan for transitioning from the construction phase into operation. A brief summary of schedules and estimated costs concludes the volume

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA

Energy Technology Data Exchange (ETDEWEB)

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1993-12-01

This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose reduction activities, with a focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and aocelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts. An author index and a subject index are provided to facilitate use. Both indices contain the abstract numbers from previous volumes, as well as the current volume. Information that the reader feels might be included in the next volume of this bibliography should be submitted to the BNL ALARA Center.

Maximal design basis accident of fusion neutron source DEMO-TIN

Energy Technology Data Exchange (ETDEWEB)

Kolbasov, B. N., E-mail: Kolbasov-BN@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

2015-12-15

When analyzing the safety of nuclear (including fusion) facilities, the maximal design basis accident at which the largest release of activity is expected must certainly be considered. Such an accident is usually the failure of cooling systems of the most thermally stressed components of a reactor (for a fusion facility, it is the divertor or the first wall). The analysis of safety of the ITER reactor and fusion power facilities (including hybrid fission–fusion facilities) shows that the initial event of such a design basis accident is a large-scale break of a pipe in the cooling system of divertor or the first wall outside the vacuum vessel of the facility. The greatest concern is caused by the possibility of hydrogen formation and the inrush of air into the vacuum chamber (VC) with the formation of a detonating mixture and a subsequent detonation explosion. To prevent such an explosion, the emergency forced termination of the fusion reaction, the mounting of shutoff valves in the cooling systems of the divertor and the first wall or blanket for reducing to a minimum the amount of water and air rushing into the VC, the injection of nitrogen or inert gas into the VC for decreasing the hydrogen and oxygen concentration, and other measures are recommended. Owing to a continuous feed-out of the molten-salt fuel mixture from the DEMO-TIN blanket with the removal period of 10 days, the radioactivity release at the accident will mainly be determined by tritium (up to 360 PBq). The activity of fission products in the facility will be up to 50 PBq.

Fusion technologies for Laser Inertial Fusion Energy (LIFE∗

Directory of Open Access Journals (Sweden)

Kramer K.J.

2013-11-01

Full Text Available The Laser Inertial Fusion-based Energy (LIFE engine design builds upon on going progress at the National Ignition Facility (NIF and offers a near-term pathway to commercial fusion. Fusion technologies that are critical to success are reflected in the design of the first wall, blanket and tritium separation subsystems. The present work describes the LIFE engine-related components and technologies. LIFE utilizes a thermally robust indirect-drive target and a chamber fill gas. Coolant selection and a large chamber solid-angle coverage provide ample tritium breeding margin and high blanket gain. Target material selection eliminates the need for aggressive chamber clearing, while enabling recycling. Demonstrated tritium separation and storage technologies limit the site tritium inventory to attractive levels. These key technologies, along with the maintenance and advanced materials qualification program have been integrated into the LIFE delivery plan. This describes the development of components and subsystems, through prototyping and integration into a First Of A Kind power plant.

First Liquid Layer Inertial Confinement Fusion Implosions at the National Ignition Facility

Science.gov (United States)

Olson, R. E.; Leeper, R. J.; Kline, J. L.; Zylstra, A. B.; Yi, S. A.; Biener, J.; Braun, T.; Kozioziemski, B. J.; Sater, J. D.; Bradley, P. A.; Peterson, R. R.; Haines, B. M.; Yin, L.; Berzak Hopkins, L. F.; Meezan, N. B.; Walters, C.; Biener, M. M.; Kong, C.; Crippen, J. W.; Kyrala, G. A.; Shah, R. C.; Herrmann, H. W.; Wilson, D. C.; Hamza, A. V.; Nikroo, A.; Batha, S. H.

2016-12-01

The first cryogenic deuterium and deuterium-tritium liquid layer implosions at the National Ignition Facility (NIF) demonstrate D2 and DT layer inertial confinement fusion (ICF) implosions that can access a low-to-moderate hot-spot convergence ratio (12 30 ) DT ice layer implosions. Although high CR is desirable in an idealized 1D sense, it amplifies the deleterious effects of asymmetries. To date, these asymmetries prevented the achievement of ignition at the NIF and are the major cause of simulation-experiment disagreement. In the initial liquid layer experiments, high neutron yields were achieved with CRs of 12-17, and the hot-spot formation is well understood, demonstrated by a good agreement between the experimental data and the radiation hydrodynamic simulations. These initial experiments open a new NIF experimental capability that provides an opportunity to explore the relationship between hot-spot convergence ratio and the robustness of hot-spot formation during ICF implosions.

Preparation for tritiated waste management of fusion facilities: Interim storage WAC

Energy Technology Data Exchange (ETDEWEB)

Decanis, C., E-mail: christelle.decanis@cea.fr [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Canas, D. [CEA, DEN/DADN, Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Derasse, F. [CEA, DEN, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Pamela, J. [CEA, Agence ITER-France, F-13108 Saint-Paul-lez-Durance (France)

2016-11-01

Highlights: • Fusion devices including ITER will generate tritiated waste. • Interim storage is the reference solution offering an answer for all types of tritiated radwaste. • Interim storage is a buffer function in the process management and definition of the waste acceptance criteria (WAC) is a key milestone in the facility development cycle. • Defining WAC is a relevant way to identify ahead of time the studies to be launched and the required actions to converge on a detailed design for example material specific studies, required treatment, interfaces management, modelling and monitoring studies. - Abstract: Considering the high mobility of tritium through the package in which it is contained, the new 50-year storage concepts proposed by the French Alternative Energies and Atomic Energy Commission (CEA) currently provide a solution adapted to the management of waste with tritium concentrations higher than the accepted limits in the disposals. The 50-year intermediate storage corresponds to 4 tritium radioactive periods i.e., a tritium reduction by a factor 16. This paper details the approach implemented to define the waste acceptance criteria (WAC) for an interim storage facility that not only takes into account the specificity of tritium provided by the reference scheme for the management of tritiated waste in France, but also the producers’ needs, the safety analysis of the facility and Andra’s disposal requirements. This will lead to define a set of waste specifications that describe the generic criteria such as acceptable waste forms, general principles and specific issues, e.g. conditioning, radioactive content, tritium content, waste tracking system, and quality control. This approach is also a way to check in advance, during the design phase of the waste treatment chain, how the future waste could be integrated into the overall waste management routes and identify possible key points that need further investigations (design changes, selection

Atomic and plasma-material interaction data for fusion. V. 6

International Nuclear Information System (INIS)

1995-01-01

Volume 6 of the supplement ''atomic and plasma-material interaction data for fusion'' to the journal ''Nuclear Fusion'' includes critical assessments and results of original experimental and theoretical studies on inelastic collision processes among the basic and dominant impurity constituents of fusion plasmas. Processes considered in the 15 papers constituting this volume are: electron impact excitation of excited Helium atoms, electron impact excitation and ionization of plasma impurity ions and atoms, electron-impurity-ion recombination and excitation, ionization and electron capture in collisions of plasma protons and impurity ions with the main fusion plasma neutral components helium and atomic and molecular hydrogen. Refs, figs, tabs

Water detritiation processing of JET purified waste water using the TRENTA facility at Tritium Laboratory Karlsruhe

Energy Technology Data Exchange (ETDEWEB)

Michling, R., E-mail: robert.michling@kit.edu; Bekris, N.; Cristescu, I.; Lohr, N.; Plusczyk, C.; Welte, S.; Wendel, J.

2013-10-15

Highlights: • Operation of a water detritiation facility under optimized conditions for high detritiation performances. • Improvement of operational procedures to process tritiated waste water. • Handling and reduction of tritiated waste water to achieve enriched low volume tritiated water for sufficient storage. • Demonstration of the efficient availability of the TRENTA WDS facility for technical scale operation. -- Abstract: A Water Detritiation System (WDS) is required for any Fusion machine in order to process tritiated waste water, which is accumulated in various subsystems during operation and maintenance. Regarding the European procurement packages for the ITER tritium fuel cycle, the WDS test facility TRENTA applying the Combined Electrolysis Catalytic Exchange (CECE) process was developed, installed and is currently in operation at the Tritium Laboratory Karlsruhe (TLK). Besides the on-going R and D work for the design of ITER WDS, the current status of the TRENTA facility provides the option to utilize the WDS for processing tritiated water. Therefore, in the framework of the EFDA JET Fusion Technology Work Programme 2011, the TLK was able to offer the capability on a representative scale to process tritiated water, which was produced during normal operation at JET. The task should demonstrate the availability of the CECE process to handle and detritiate the water in terms of tritium enrichment and volume reduction. The operational program comprised the processing of purified tritiated water from JET, with a total volume of 180 l and an activity of 74 GBq. The paper will give an introduction to the TRENTA WDS facility and an overview of the operational procedure regarding tritiated water reduction. Data concerning required operation time, decontamination and enrichment performances and different operating procedures will be presented as well. Finally, a preliminary study on a technical implementation of processing the entire stock of JET

Water detritiation processing of JET purified waste water using the TRENTA facility at Tritium Laboratory Karlsruhe

International Nuclear Information System (INIS)

Michling, R.; Bekris, N.; Cristescu, I.; Lohr, N.; Plusczyk, C.; Welte, S.; Wendel, J.

2013-01-01

Highlights: • Operation of a water detritiation facility under optimized conditions for high detritiation performances. • Improvement of operational procedures to process tritiated waste water. • Handling and reduction of tritiated waste water to achieve enriched low volume tritiated water for sufficient storage. • Demonstration of the efficient availability of the TRENTA WDS facility for technical scale operation. -- Abstract: A Water Detritiation System (WDS) is required for any Fusion machine in order to process tritiated waste water, which is accumulated in various subsystems during operation and maintenance. Regarding the European procurement packages for the ITER tritium fuel cycle, the WDS test facility TRENTA applying the Combined Electrolysis Catalytic Exchange (CECE) process was developed, installed and is currently in operation at the Tritium Laboratory Karlsruhe (TLK). Besides the on-going R and D work for the design of ITER WDS, the current status of the TRENTA facility provides the option to utilize the WDS for processing tritiated water. Therefore, in the framework of the EFDA JET Fusion Technology Work Programme 2011, the TLK was able to offer the capability on a representative scale to process tritiated water, which was produced during normal operation at JET. The task should demonstrate the availability of the CECE process to handle and detritiate the water in terms of tritium enrichment and volume reduction. The operational program comprised the processing of purified tritiated water from JET, with a total volume of 180 l and an activity of 74 GBq. The paper will give an introduction to the TRENTA WDS facility and an overview of the operational procedure regarding tritiated water reduction. Data concerning required operation time, decontamination and enrichment performances and different operating procedures will be presented as well. Finally, a preliminary study on a technical implementation of processing the entire stock of JET

High concentration tritium gas measurement with small volume ionization chambers for fusion fuel gas monitors

International Nuclear Information System (INIS)

Uda, Tatsuhiko; Okuno, Kenji; Matsuda, Yuji; Naruse, Yuji

1991-01-01

To apply ionization chambers to fusion fuel gas processing systems, high concentration tritium gas was experimentally measured with small volume 0.16 and 21.6 cm 3 ionization chambers. From plateau curves, the optimum electric field strength was obtained as 100∼200 V/cm. Detection efficiency was confirmed as dependent on the ionization ability of the filled gas, and moreover on its stopping power, because when the range of the β-rays was shortened, the probability of energy loss by collisions with the electrode and chamber wall increased. Loss of ions by recombination was prevented by using a small volume ionization chamber. For example the 0.16 cm 3 ionization chamber gave measurement with linearity to above 40% tritium gas. After the tritium gas measurements, the concentration levels inside the chamber were estimated from their memory currents. Although more than 1/4,000 of the maximum, current was observed as a memory effect, the smaller ionization chamber gave a smaller memory effect. (author)

Fusion the energy of the universe

CERN Document Server

McCracken, Garry

2012-01-01

Fusion: The Energy of the Universe, 2e is an essential reference providing basic principles of fusion energy from its history to the issues and realities progressing from the present day energy crisis. The book provides detailed developments and applications for researchers entering the field of fusion energy research. This second edition includes the latest results from the National Ignition Facility at the Lawrence Radiation Laboratory at Livermore, CA, and the progress on the International Thermonuclear Experimental Reactor (ITER) tokamak programme at Caderache, France.

Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

Energy Technology Data Exchange (ETDEWEB)

1984-01-01

This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

International Nuclear Information System (INIS)

1984-01-01

This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

Information base for waste repository design. Volume 5. Decommissioning of underground facilities. Technical report

International Nuclear Information System (INIS)

Giuffre, M.S.; Plum, R.L.; Koplik, C.M.; Talbot, R.

1979-03-01

This report is Volume 5 of a seven volume document on nuclear waste repository design issues. This report discusses the requirements for decommissioning a deep underground facility for the disposal of radioactive waste. The techniques for sealing the mined excavations are presented and an information base on potential backfill materials is provided. Possible requirements for monitoring the site are discussed. The performance requirements for backfill materials are outlined. The advantages and disadvantages of each sealing method are stated

In pursuit of fusion; ARGUS laser system at Livermore

International Nuclear Information System (INIS)

Simmons, W.W.

1976-01-01

The ARGUS laser facility has been developed to achieve significant laser fusion milestones; high density (greater than 10 g/cm 3 ) implosions, high temperature (greater than 10 KeV) implosions, and high yield from advanced target designs. The ARGUS laser, central to this facility is a twin-beam, 20 cm output aperture, Nd:glass solid state laser capable of delivering greater than 3 TW of power to laser fusion targets. At the present time, ARGUS is fully operational, and has produced up to 10 9 neutrons in selected target irradiation experiments. The performance of this facility is described

Fusion Safety Program Annual Report, Fiscal Year 1996

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Cadwallader, L.C.

1996-12-01

This report summarizes the major activities of the Fusion Safety Program in FY 1996. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory, and Lockheed Martin Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. The objective is to perform research and develop data needed to ensure safety in fusion facilities. Activities include experiments, analysis, code development and application, and other forms of research. These activities are conducted at the INEL, at other DOE laboratories, and at other institutions. Among the technical areas covered in this report are tritium safety, chemical reactions and activation product release, risk assessment failure rate database development, and safety code development and application to fusion safety issues. Most of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Work done for ITER this year has focused on developing the needed information for the Non- Site- Specific Safety Report (NSSR-1). A final area of activity described is development of the new DOE Technical Standards for Safety of Magnetic Fusion Facilities

DOE Handbook: Supplementary guidance and design experience for the fusion safety standards DOE-STD-6002-96 and DOE-STD-6003-96

Energy Technology Data Exchange (ETDEWEB)

None

1999-01-01

Two standards have been developed that pertain to the safety of fusion facilities. These are DOE- STD-6002-96, Safety of Magnetic Fusion Facilities: Requirements, and DOE-STD-6003-96, Safety of Magnetic Fusion Facilities: Guidance. The first of these standards identifies requirements that subscribers to that standard must meet to achieve safety in fusion facilities. The second standard contains guidance to assist in meeting the requirements identified in the first This handbook provides additional documentation on good operations and design practices as well as lessons learned from the experiences of designers and operators of previous fusion facilities and related systems. It is intended to capture the experience gained in the various fields and pass it on to designers of future fusion facilities as a means of enhancing success and safety. The sections of this document are presented according to the physical location of the major systems of a fusion facility, beginning with the vacuum vessel and proceeding to those systems and components outside the vacuum vessel (the "Ex-vessel Systems"). The last section describes administrative procedures that cannot be localized to specific components. It has been tacitly assumed that the general structure of the fusion facilities addressed is that of a tokamak though the same principles would apply to other magnetic confinement options.

Argus Laser Fusion Facility

International Nuclear Information System (INIS)

Speck, D.R.; Simmons, W.W.

1976-01-01

ARGUS is a two-beam Nd: glass laser system built for laser fusion irradiation experiments. It is the first glass laser system planned and built with the understanding that small-scale beam break-up is the dominant performance limiting factor in obtaining high output power. Accordingly, five vacuum spatial filters are located at strategic intervals along each chain to eliminate the accumulated small-scale filamentation. This strategy permits cascading of amplifiers to obtain a focusable output of more than one terawatt per arm in a spatially clean beam of 20 centimeter diameter. Beam diagnostics which characterize each shot include the time-integrated spatial profile and the time resolved intensity/power at the target. Demonstrated performance to date includes: (1) Peak power in excess of 2 TW at the target is achieved with regularity. (2) Maximum system brightness is in excess of 10 17 watts/cm 2 ster. (3) Shot-to-shot pointing stability within 50 μ radians is achieved over periods of days. (4) Successful target experiments have been performed with pulses of from 30 to 500 ps duration

Incomplete fusion studies

International Nuclear Information System (INIS)

Singh, B.P.

2011-01-01

In order to study the incomplete fusion reaction dynamics at energies ≅ 4-7 MeV/nucleon, several experiments have been carried out using accelerator facilities available in India. The measurements presented here cover a wide range of projectile-target combinations and enhance significantly our knowledge about incomplete fusion reaction dynamics. Here, the three sets of measurements have been presented; (i) excitation functions, (ii) forward recoil range distributions and (iii) the spin distributions. The first evidence of these reactions has been obtained from the measurement and analysis of excitation functions for xn/αxn/2αxn-channels. The measured excitation functions have been analyzed within the framework of compound nucleus model. The results obtained indicate the occurrence of fusion incompleteness at low beam energies. However, in order to determine the relative contribution of complete and incomplete fusion reaction processes, the recoil range distributions of the heavy residues have also been measured and analyzed within the framework of breakup fusion model which confirmed the fusion incompleteness in several heavy ion reactions involving α-emitting reaction channels. Further, in order to study the role of l-values in these reactions the spin distributions of the residues populated in case of complete and incomplete channels have been measured and are found to be distinctly different. The analysis of the data on spin distribution measurements indicate that the mean values of driving input angular momenta associated with direct-α-emitting (incomplete fusion) channels are higher than that observed for fusion-evaporation xn or α-emitting (complete fusion) channels, and is found to increase with direct α-multiplicity in the forward cone. One of the important conclusions drawn in the present work is that, there is significant incomplete fusion contribution even at energies slightly above the barrier. Further, the projectile structure has been found to

FMIT: an accelerator-based neutron factory for fusion materials qualification

International Nuclear Information System (INIS)

Burke, R.J.; Hagan, J.W.; Trego, A.L.

1983-01-01

The Fusion Materials Irradiation Test Facility will provide a unique testing environment for irradiation of structural and special-purpose materials in support of fusion-power systems. The neutron source will be produced by a deuteron-lithium stripping reaction to generate high-energy neutrons to ensure materials damage characteristic of the deuterium-tritium power system. The facility, its testing role, the status, and major aspects of its design and supporting system development are described. Emphasis is given to programmatic elements and features incorporated in the accelerator and other systems to assure that the FMIT runs as a highly reliable fusion materials testing installation

Thermonuclear controlled fusion: international cooperation

International Nuclear Information System (INIS)

Conscience, J.-F.

2001-01-01

This report summarizes the current worldwide status of research in the field of thermonuclear controlled fusion as well as the international research programme planed for the next decades. The two main projects will be the ITER facility (International Thermonuclear Experimental Reactor) that should produce 10 times more energy than the energy injected, and the IFMIF (International Fusion Materials Irradiation Facility) designed to study the reactions of materials under intense neutron fluxes. The future of the pioneering JET facility (Joint European Torus) is also discussed. The engagement of the various countries (USA, Japan, Germany, Russian Federation and Canada) and international organisations (EURATOM and IEA) in terms of investment and research is described. Switzerland is involved in this program through an agreement with EURATOM and is mainly dedicated to experimental studies with the TCV machine in Lausanne and numerical studies of plasma configurations. It will participate to the development of the microwave plasma heating system for the ITER machine

Chemical Hazards and Safety Issues in Fusion Safety Design

International Nuclear Information System (INIS)

Cadwallader, L.C.

2003-01-01

Radiological inventory releases have dominated accident consequences for fusion; these consequences are important to analyze and are generally the most severe result of a fusion facility accident event. However, the advent of, or plan for, large-scale usage of some toxic materials poses the additional hazard of chemical exposure from an accident event. Examples of toxic chemicals are beryllium for magnetic fusion and fluorine for laser fusion. Therefore, chemical exposure consequences must also be addressed in fusion safety assessment. This paper provides guidance for fusion safety analysis. US Department of Energy (DOE) chemical safety assessment practices for workers and the public are reviewed. The US Environmental Protection Agency (EPA) has published some guidance on public exposure to releases of mixtures of chemicals, this guidance has been used to create an initial guideline for treating mixed radiological and toxicological releases in fusion; for example, tritiated hazardous dust from a tokamak vacuum vessel. There is no convenient means to judge the hazard severity of exposure to mixed materials. The chemical fate of mixed material constituents must be reviewed to determine if there is a separate or combined radiological and toxicological carcinogenesis, or if other health threats exist with radiological carcinogenesis. Recommendations are made for fusion facility chemical safety evaluation and safety guidance for protecting the public from chemical releases, since such levels are not specifically identified in the DOE fusion safety standard

Physics of laser fusion. Volume III. High-power pulsed lasers

International Nuclear Information System (INIS)

Holzrichter, J.F.; Eimerl, D.; George, E.V.; Trenholme, J.B.; Simmons, W.W.; Hunt, J.T.

1982-09-01

High-power pulsed lasers can deliver sufficient energy on inertial-confinement fusion (ICF) time scales (0.1 to 10 ns) to heat and compress deuterium-tritium fuel to fusion-reaction conditions. Several laser systems have been examined, including Nd:glass, CO 2 , KrF, and I 2 , for their ICF applicability. A great deal of developmental effort has been applied to the Nd:glass laser and the CO 2 gas laser systems; these systems now deliver > 10 4 J and 20 x 10 12 W to ICF targets. We are constructing the Nova Nd:glass laser at LLNL to provide > 100 kJ and > 100 x 10 12 W of 1-μm radiation for fusion experimentation in the mid-1980s. For ICF target gain > 100 times the laser input, we expect that the laser driver must deliver approx. 3 to 5 MJ of energy on a time scale of 10 to 20 ns. In this paper we review the technological status of fusion-laser systems and outline approaches to constructing high-power pulsed laser drivers

STAR facility tritium accountancy

International Nuclear Information System (INIS)

Pawelko, R. J.; Sharpe, J. P.; Denny, B. J.

2008-01-01

The Safety and Tritium Applied Research (STAR) facility has been established to provide a laboratory infrastructure for the fusion community to study tritium science associated with the development of safe fusion energy and other technologies. STAR is a radiological facility with an administrative total tritium inventory limit of 1.5 g (14,429 Ci) [1]. Research studies with moderate tritium quantities and various radionuclides are performed in STAR. Successful operation of the STAR facility requires the ability to receive, inventory, store, dispense tritium to experiments, and to dispose of tritiated waste while accurately monitoring the tritium inventory in the facility. This paper describes tritium accountancy in the STAR facility. A primary accountancy instrument is the tritium Storage and Assay System (SAS): a system designed to receive, assay, store, and dispense tritium to experiments. Presented are the methods used to calibrate and operate the SAS. Accountancy processes utilizing the Tritium Cleanup System (TCS), and the Stack Tritium Monitoring System (STMS) are also discussed. Also presented are the equations used to quantify the amount of tritium being received into the facility, transferred to experiments, and removed from the facility. Finally, the STAR tritium accountability database is discussed. (authors)

Fusion at counterstreaming ion beams - ion optic fusion (IOF)

International Nuclear Information System (INIS)

Gryzinski, M.

1981-01-01

The results of investigation are briefly reviewed in the field of ion optic fusion performed at the Institute of Nuclear Research in Swierk. The ion optic fusion concept is based on the possibility of obtaining fusion energy at highly ordered motion of ions in counterstreaming ion beams. For this purpose TW ion beams must be produced and focused. To produce dense and charge-neutralized ion beams the selective conductivity and ballistic focusing ideas were formulated and used in a series of RPI devices with low-pressure cylindrical discharge between grid-type electrodes. 100 kA, 30 keV deuteron beams were successfully produced and focused into the volume of 1 cm 3 , yielding 10 9 neutrons per 200 ns shot on a heavy ice target. Cylindrically convergent ion beams with magnetic anti-defocusing were proposed in order to reach a positive energy gain at reasonable energy level. (J.U.)

Data security on the national fusion grid

Energy Technology Data Exchange (ETDEWEB)

Burruss, Justine R.; Fredian, Tom W.; Thompson, Mary R.

2005-06-01

The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER.

Data security on the national fusion grid

International Nuclear Information System (INIS)

Burruss, Justine R.; Fredian, Tom W.; Thompson, Mary R.

2005-01-01

The National Fusion Collaboratory project is developing and deploying new distributed computing and remote collaboration technologies with the goal of advancing magnetic fusion energy research. This work has led to the development of the US Fusion Grid (FusionGrid), a computational grid composed of collaborative, compute, and data resources from the three large US fusion research facilities and with users both in the US and in Europe. Critical to the development of FusionGrid was the creation and deployment of technologies to ensure security in a heterogeneous environment. These solutions to the problems of authentication, authorization, data transfer, and secure data storage, as well as the lessons learned during the development of these solutions, may be applied outside of FusionGrid and scale to future computing infrastructures such as those for next-generation devices like ITER

FENIX [Fusion ENgineering International eXperimental]: A test facility for ITER [International Thermonuclear Experimental Reactor] and other new superconducting magnets

International Nuclear Information System (INIS)

Slack, D.S.; Patrick, R.E.; Miller, J.R.

1990-01-01

The Fusion ENgineering International eXperimental (FENIX) Test Facility which is nearing completion at Lawrence Livermore National Laboratory, is a 76-t set of superconducting magnets housed in a 4-m-diameter cryostat. It represents a significant step toward meeting the testing needs for the development of superconductors appropriate for large-scale magnet applications such as the International Thermonuclear Experimental Reactor (ITER). The magnet set is configured to allow radial access to the 0.4-m-diameter high-field region where maximum fields up to 14 T will be provided. The facility is fitted with a thermally isolated test well with a port to the high-field region that allows insertion and removal of test conductors without disturbing the cryogenic environment of the magnets. It is expected that the facility will be made available to magnet developers internationally, and this paper discusses its general design features, its construction, and its capabilities

IFMIF, a fusion relevant neutron source for material irradiation current status

International Nuclear Information System (INIS)

Knaster, J.; Chel, S.; Fischer, U.; Groeschel, F.; Heidinger, R.; Ibarra, A.; Micciche, G.; Möslang, A.; Sugimoto, M.; Wakai, E.

2014-01-01

The d-Li based International Fusion Materials Irradiation Facility (IFMIF) will provide a high neutron intensity neutron source with a suitable neutron spectrum to fulfil the requirements for testing and qualifying fusion materials under fusion reactor relevant irradiation conditions. The IFMIF project, presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the Broader Approach (BA) Agreement between Japan Government and EURATOM, aims at the construction and testing of the most challenging facility sub-systems, such as the first accelerator stage, the Li target and loop, and irradiation test modules, as well as the design of the entire facility, thus to be ready for the IFMIF construction with a clear understanding of schedule and cost at the termination of the BA mid-2017. The paper reviews the IFMIF facility and its principles, and reports on the status of the EVEDA activities and achievements

Health physics in fusion reactor design

International Nuclear Information System (INIS)

Wong, K.Y.; Dinner, P.J.

1984-06-01

Experience in the control of tritium exposures to workers and the public gained through the design and operation of Ontario Hydro's nuclear stations has been applied to fusion projects and to design studies on emerging fusion reactor concepts. Ontario Hydro performance in occupational tritium exposure control and environmental impact is reviewed. Application of tritium control technologies and dose management methodology during facility design is highlighted

Ocean Thermal Energy Converstion (OTEC) test facilities study program. Final report. Volume II. Part B

Energy Technology Data Exchange (ETDEWEB)

None

1977-01-17

Results are presented of an 8-month study to develop alternative non-site-specific OTEC facilities/platform requirements for an integrated OTEC test program which may include land and floating test facilities. Volume II--Appendixes is bound in three parts (A, B, and C) which together comprise a compendium of the most significant detailed data developed during the study. Part B provides an annotated test list and describes component tests and system tests.

The heavy ion fusion program in the USA

International Nuclear Information System (INIS)

Bangerter, R.O.; Davidson, R.C.; Herrmannsfeldt, W.B.; Lindl, J.; Meier, W.R.; Logan, B.G.

2001-01-01

Inertial fusion energy research has enjoyed increased interest and funding. This has allowed expanded programs in target design, target fabrication, fusion chamber research, target injection and tracking, and accelerator research. The target design effort examines ways to minimize the beam power and energy and increase the allowable focal spot size while preserving target gain. Chamber research for heavy ion fusion emphasizes the use of thick liquid walls to serve as the coolant, breed tritium, and protect the structural wall from neutrons, photons, and other target products. Several small facilities are now operating to model fluid chamber dynamics. A facility to study target injection and tracking has been built and a second facility is being designed. Improved economics is an important goal of the accelerator research. The accelerator research is also directed toward the design of an Integrated Research Experiment (IRE). The IRE is being designed to accelerate ions to >100 MeV, enabling experiments in beam dynamics, focusing, and target physics. Activities leading to the IRE include ion source development and a High Current Experiment (HCX) designed to transport and accelerate a single beam of ions with a beam current of approximately 1 A, the initial current required for each beam of a fusion driver. In terms of theory, the program is developing a source-to-target numerical simulation capability. The goal of the entire program is to enable an informed decision about the promise of heavy ion fusion in about a decade. (author)

Utilities and offsites design baseline. Outside Battery Limits Facility 6000 tpd SRC-I Demonstration Plant. Volume 1

Energy Technology Data Exchange (ETDEWEB)

None

1984-05-25

As part of the overall Solvent Refined Coal (SRC-1) project baseline being prepared by International Coal Refining Company (ICRC), the RUST Engineering Company is providing necessary input for the Outside Battery Limits (OSBL) Facilities. The project baseline is comprised of: design baseline - technical definition of work; schedule baseline - detailed and management level 1 schedules; and cost baseline - estimates and cost/manpower plan. The design baseline (technical definition) for the OSBL Facilities has been completed and is presented in Volumes I, II, III, IV, V and VI. The OSBL technical definition is based on, and compatible with, the ICRC defined statement of work, design basis memorandum, master project procedures, process and mechanical design criteria, and baseline guidance documents. The design basis memorandum is included in Paragraph 1.3 of Volume I. The baseline design data is presented in 6 volumes. Volume I contains the introduction section and utility systems data through steam and feedwater. Volume II continues with utility systems data through fuel system, and contains the interconnecting systems and utility system integration information. Volume III contains the offsites data through water and waste treatment. Volume IV continues with offsites data, including site development and buildings, and contains raw materials and product handling and storage information. Volume V contains wastewater treatment and solid wastes landfill systems developed by Catalytic, Inc. to supplement the information contained in Volume III. Volume VI contains proprietary information of Resources Conservation Company related to the evaporator/crystallizer system of the wastewater treatment area.

Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

Energy Technology Data Exchange (ETDEWEB)

1982-09-01

This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

International Nuclear Information System (INIS)

1982-09-01

This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

Diagnostic measurements related to laser driven inertial confinement fusion

International Nuclear Information System (INIS)

Campbell, D.E.

1979-01-01

Scientists at the Lawrence Livermore Laboratory have been conducting laser driven inertial confinement fusion experiments for over five years. The first proof of the thermonuclear burn came at the Janus target irradiation facility in the spring of 1975. Since that time three succeedingly higher energy facilities have been constructed at Livermore, Cyclops, Argus and Shiva, where increased fusion efficiency has been demonstrated. A new facility, called Nova, is now in the construction phase and we are hopeful that scientific break even (energy released compared to incident laser energy on target) will be demonstrated here in early 1980's. Projected progress of the Livermore program is shown

Oak Ridge rf Test Facility

International Nuclear Information System (INIS)

Gardner, W.L.; Hoffman, D.J.; McCurdy, H.C.; McManamy, T.J.; Moeller, J.A.; Ryan, P.M.

1985-01-01

The rf Test Facility (RFTF) of Oak Ridge National Laboratory (ORNL) provides a national facility for the testing and evaluation of steady-state, high-power (approx.1.0-MW) ion cyclotron resonance heating (ICRH) systems and components. The facility consists of a vacuum vessel and two fully tested superconducting development magnets from the ELMO Bumpy Torus Proof-of-Principle (EBT-P) program. These are arranged as a simple mirror with a mirror ratio of 4.8. The axial centerline distance between magnet throat centers is 112 cm. The vacuum vessel cavity has a large port (74 by 163 cm) and a test volume adequate for testing prototypic launchers for Doublet III-D (DIII-D), Tore Supra, and the Tokamak Fusion Test Reactor (TFTR). Attached to the internal vessel walls are water-cooled panels for removing the injected rf power. The magnets are capable of generating a steady-state field of approx.3 T on axis in the magnet throats. Steady-state plasmas are generated in the facility by cyclotron resonance breakdown using a dedicated 200-kW, 28-GHz gyrotron. Available rf sources cover a frequency range of 2 to 200 MHz at 1.5 kW and 3 to 18 MHz at 200 kW, with several sources at intermediate parameters. Available in July 1986 will be a >1.0-MW, cw source spanning 40 to 80 MHz. 5 figs

Preliminary results of the International Fusion Materials Irradiation Facility deuteron injector

Energy Technology Data Exchange (ETDEWEB)

Gobin, R.; Adroit, G.; Bogard, D.; Bourdelle, G.; Chauvin, N.; Delferriere, O.; Gauthier, Y.; Girardot, P.; Guiho, P.; Harrault, F.; Jannin, J. L.; Loiseau, D.; Mattei, P.; Roger, A.; Sauce, Y.; Senee, F.; Vacher, T. [Commissariat a l' Energie Atomique et aux Energie Alternatives, CEA/Saclay, DSM/IRFU, 91191-Gif/Yvette (France)

2012-02-15

In the framework of the IFMIF-EVEDA project (International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities), CEA/IRFU is in charge of the design, construction, and characterization of the 140 mA continuous deuteron injector, including the source and the low energy beam line. The electron cyclotron resonance ion source which operates at 2.45 GHz is associated with a 4-electrode extraction system in order to minimize beam divergence at the source exit. Krypton gas injection is foreseen in the 2-solenoid low energy beam line. Such Kr injection will allow reaching a high level of space charge compensation in order to improve the beam matching at the radio frequency quadrupole (RFQ) entrance. The injector construction is now completed on the Saclay site and the first plasma and beam production has been produced in May 2011. This installation will be tested with proton and deuteron beams either in pulsed or continuous mode at Saclay before shipping to Japan. In this paper, after a brief description of the installation, the preliminary results obtained with hydrogen gas injection into the plasma chamber will be reported.

DOE Handbook: Supplementary guidance and design experience for the fusion safety standards DOE-STD-6002-96 and DOE-STD-6003-96

International Nuclear Information System (INIS)

1999-01-01

Two standards have been developed that pertain to the safety of fusion facilities. These are DOE- STD-6002-96, Safety of Magnetic Fusion Facilities: Requirements, and DOE-STD-6003-96, Safety of Magnetic Fusion Facilities: Guidance. The first of these standards identifies requirements that subscribers to that standard must meet to achieve safety in fusion facilities. The second standard contains guidance to assist in meeting the requirements identified inthefirst This handbook provides additional documentation on good operations and design practices as well as lessons learned from the experiences of designers and operators of previous fusion facilities and related systems. It is intended to capture the experience gained in the various fields and pass it on to designers of future fusion facilities as a means of enhancing success and safeiy. The sections of this document are presented according to the physical location of the major systems of a t%sion facility, beginning with the vacuum vessel and proceeding to those systems and components outside the vacuum vessel (the ''Ex-vessel Systems''). The last section describes administrative procedures that cannot be localized to specific components. It has been tacitly assumed that the general structure of the fusion facilities addressed is that of a tokamak though the same principles would apply to other magnetic confinement options

Waste Receiving and Processing Facility Module 2A: Advanced Conceptual Design Report. Volume 1

Energy Technology Data Exchange (ETDEWEB)

1994-03-01

This ACDR was performed following completed of the Conceptual Design Report in July 1992; the work encompassed August 1992 to January 1994. Mission of the WRAP Module 2A facility is to receive, process, package, certify, and ship for permanent burial at the Hanford site disposal facilities the Category 1 and 3 contact handled low-level radioactive mixed wastes that are currently in retrievable storage at Hanford and are forecast to be generated over the next 30 years by Hanford, and waste to be shipped to Hanford from about DOE sites. This volume provides an introduction to the ACDR process and the scope of the task along with a project summary of the facility, treatment technologies, cost, and schedule. Major areas of departure from the CDR are highlighted. Descriptions of the facility layout and operations are included.

The National Ignition Facility Project. Revision 1

International Nuclear Information System (INIS)

Paisner, J.A.; Campbell, E.M.; Hogan, W.J.

1994-01-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. This paper reviews the design, schedule, and costs associated with the construction project

Elements of Successful and Safe Fusion Experiment Operations

International Nuclear Information System (INIS)

Rule, K.; Cadwallader, L.; Takase, Y.; Norimatsu, T.; Kaneko, O.; Sato, M.; Savercool, R.

2009-01-01

A group of fusion safety professionals contribute to a Joint Working Group (JWG) that performs occupational safety walkthroughs of US and Japanese fusion experiments on a routine basis to enhance the safety of visiting researchers. The most recent walkthrough was completed in Japan in March 2008 by the US Safety Monitor team. This paper gives the general conclusions on fusion facility personnel safety that can be drawn from the series of walkthroughs

Overview of US fast-neutron facilities and testing capabilities

International Nuclear Information System (INIS)

Evans, E.A.; Cox, C.M.; Jackson, R.J.

1982-01-01

Rather than attempt a cataloging of the various fast neutron facilities developed and used in this country over the last 30 years, this paper will focus on those facilities which have been used to develop, proof test, and explore safety issues of fuels, materials and components for the breeder and fusion program. This survey paper will attempt to relate the evolution of facility capabilities with the evolution of development program which use the facilities. The work horse facilities for the breeder program are EBR-II, FFTF and TREAT. For the fusion program, RTNS-II and FMIT were selected

Hybrid fission-fusion nuclear reactors

International Nuclear Information System (INIS)

Zucchetti, Massimo

2011-01-01

A fusion-fission hybrid could contribute to all components of nuclear power - fuel supply, electricity production, and waste management. The idea of the fusion-fission hybrid is many decades old. Several ideas, both new and revisited, have been investigated by hybrid proponents. These ideas appear to have attractive features, but they require various levels of advances in plasma science and fusion and nuclear technology. As a first step towards the development of hybrid reactors, fusion neutron sources can be considered as an option. Compact high-field tokamaks can be a candidate for being the neutron source in a fission-fusion hybrid, essentially due to their design characteristics, such as compact dimensions, high magnetic field, flexibility of operation. This study presents the development of a tokamak neutron source for a material testing facility using an Ignitor-based concept. The computed values show the potential of this neutron-rich device for fusion materials testing. Some full-power months of operation are sufficient to obtain relevant radiation damage values in terms of dpa. (Author)

Atomic and plasma-material interaction data for fusion. V. 3

International Nuclear Information System (INIS)

1992-01-01

This volume of Atomic and Plasma-Material Interaction Data for Fusion is devoted to atomic collision processes of helium atoms and of beryllium and boron atoms and ions in fusion plasmas. Most of the articles included in this volume are extended versions of the contributions presented at the IAEA experts' meetings on Atomic Data for Helium Beam Fusion Alpha Particle Diagnostics and on the Atomic Database for Beryllium and Boron, held in June 1991 at the IAEA headquarters in Vienna, or have resulted from the cross-section data analyses and evaluations performed by the working groups of these meetings. Refs, figs and tabs

Accelerator and Fusion Research Division: Summary of activities, 1986

International Nuclear Information System (INIS)

1987-01-01

This report contains a summary of activities at the Lawrence Berkeley Laboratory's Accelerator and Fusion Research Division for the year 1986. Topics and facilities investigated in individual papers are: 1-2 GeV Synchrotron Radiation Source, the Center for X-Ray Optics, Accelerator Operations, High-Energy Physics Technology, Heavy-Ion Fusion Accelerator Research and Magnetic Fusion Energy. Six individual papers have been indexed separately

EDITORIAL: Safety aspects of fusion power plants

Science.gov (United States)

Kolbasov, B. N.

2007-07-01

This special issue of Nuclear Fusion contains 13 informative papers that were initially presented at the 8th IAEA Technical Meeting on Fusion Power Plant Safety held in Vienna, Austria, 10-13 July 2006. Following recommendation from the International Fusion Research Council, the IAEA organizes Technical Meetings on Fusion Safety with the aim to bring together experts to discuss the ongoing work, share new ideas and outline general guidance and recommendations on different issues related to safety and environmental (S&E) aspects of fusion research and power facilities. Previous meetings in this series were held in Vienna, Austria (1980), Ispra, Italy (1983), Culham, UK (1986), Jackson Hole, USA (1989), Toronto, Canada (1993), Naka, Japan (1996) and Cannes, France (2000). The recognized progress in fusion research and technology over the last quarter of a century has boosted the awareness of the potential of fusion to be a practically inexhaustible and clean source of energy. The decision to construct the International Thermonuclear Experimental Reactor (ITER) represents a landmark in the path to fusion power engineering. Ongoing activities to license ITER in France look for an adequate balance between technological and scientific deliverables and complying with safety requirements. Actually, this is the first instance of licensing a representative fusion machine, and it will very likely shape the way in which a more common basis for establishing safety standards and policies for licensing future fusion power plants will be developed. Now that ITER licensing activities are underway, it is becoming clear that the international fusion community should strengthen its efforts in the area of designing the next generations of fusion power plants—demonstrational and commercial. Therefore, the 8th IAEA Technical Meeting on Fusion Safety focused on the safety aspects of power facilities. Some ITER-related safety issues were reported and discussed owing to their potential

Investigation of high flux test module for the international fusion materials irradiation facilities (IFMIF)

International Nuclear Information System (INIS)

Miyashita, Makoto; Sugimoto, Masayoshi; Yutani, Toshiaki

2007-03-01

This report describes investigation on structure of a high neutron flux test module (HFTM) for the International Fusion Materials Irradiation Facilities (IFMIF). The HFTM is aimed for neutron irradiation of a specimen in a high neutron flux domain of the test cell for irradiation ground of IFMIF. We investigated the overall structure of the HFTM that was able to include specimens in a rig and thermocouple arrangement, an interface of control signal and support structure. Moreover, pressure and the amount of the bend in the module vessel (a rectangular section pressure vessel) were calculated. The module vessel did a rectangular section from limitation of a high neutron flux domain. Also, we investigated damage of thermocouples under neutron irradiation, which was a temperature sensor of irradiation materials temperature control demanded high precision. Based on these results, drawings on the HTFM structure. (author)

Fuel conditioning facility electrorefiner volume calibration

International Nuclear Information System (INIS)

Bucher, R.G.; Orechwa, Y.

1995-01-01

In one of the electrometallurgical process steps of the Fuel Conditioning Facility (FCF), die in-process nuclear material is dissolved in the electrorefiner tank in an upper layer of a mixture of liquid LiCl-KCl salt and a lower layer of liquid cadmium. The electrorefiner tank, as most process tanks, is not a smooth right-circular cylinder for which a single linear volume calibration curve could be fitted over the whole height of the tank. Rather, the tank contains many internal components, which cause systematic deviations from a single linear function. The nominal operating temperature of the electrorefiner is 500 degrees C although the salt and cadmium are introduced at 410 degrees C. The operating materials and temperatures preclude multiple calibration runs at operating conditions. In order to maximize the calibration information, multiple calibration runs were performed with water at room temperature. These data allow identification of calibration segments, and preliminary estimation of the calibration function and calibration uncertainties. The final calibration function is based on a combination of data from die water calibrations and the measurements made during the filling of the electrorefiner with salt and cadmium for operation

FINESSE: study of the issues, experiments and facilities for fusion nuclear technology research and development. Interim report. Volume III

Energy Technology Data Exchange (ETDEWEB)

Abdou, M.

1984-10-01

This chapter deals with the analysis and engineering scaling of solid breeded blankets. The limits under which full component behavior can be achieved under changed test conditions are explored. The characterization of these test requirements for integrated testing contributes to the overall test matrix and test plan for the understanding and development of fusion nuclear technology. The second chapter covers the analysis and engineering scaling of liquid metal blankets. The testing goals for a complete blanket program are described. (MOW)

Modeling, analysis and experiments for fusion nuclear technology

International Nuclear Information System (INIS)

Abdou, M.A.; Hadid, A.H.; Raffray, A.R.; Tillack, M.S.; Iizuka, T.

1988-01-01

Selected issues in the development of fusion nuclear technology (FNT) have been studied. These relate to (1) near-term experiments, modeling, and analysis for several key FNT issues, and (2) FNT testing in future fusion facilities. A key concern for solid breeder blankets is to reduce the number of candidate materials and configurations for advanced experiments to emphasize those with the highest potential. Based on technical analysis, recommendations have been developed for reducing the size of the test matrix and for focusing the testing program on important areas of emphasis. The characteristics of an advanced liquid metal MHD experiment have also been studied. This facility is required in addition to existing facilities in order to address critical uncertainties in MHD fluid flow and heat transfer. In addition to experiments, successful development of FNT will require models for interpreting experimental data, for planning experiments, and for use as a design tool for fusion components. Modeling of liquid metal fluid flows is a particular area of need in which substantial progress is expected, and initial efforts are reported here. Preliminary results on the modeling of tritium transport and inventory in solid breeders are also summarized. Finally, the thermo-mechanical behavior of liquid-metal-cooled limiters is analyzed and the parameter space for feasible designs is explored. Because of the renewed strong interest in a fusion engineering facility, a critical review and analysis of the important FNT testing requirements have been performed. Several areas have been emphasized due to their strong impact on the design and cost of the test facility. These include (1) the length of the plasma burn and the mode of operation (pulsed vs. steady-state), and (2) the need for a tritium-producing blanket and its impact on the availability of the device. (orig.)

Ventilation Systems Operating Experience Review for Fusion Applications

International Nuclear Information System (INIS)

Cadwallader, L.C.

1999-01-01

This report is a collection and review of system operation and failure experiences for air ventilation systems in nuclear facilities. These experiences are applicable for magnetic and inertial fusion facilities since air ventilation systems are support systems that can be considered generic to nuclear facilities. The report contains descriptions of ventilation system components, operating experiences with these systems, component failure rates, and component repair times. Since ventilation systems have a role in mitigating accident releases in nuclear facilities, these data are useful in safety analysis and risk assessment of public safety. An effort has also been given to identifying any safety issues with personnel operating or maintaining ventilation systems. Finally, the recommended failure data were compared to an independent data set to determine the accuracy of individual values. This comparison is useful for the International Energy Agency task on fusion component failure rate data collection

Construction and testing of the Mirror Fusion Test Facility magnets

International Nuclear Information System (INIS)

Kozman, T.; Shimer, D.; VanSant, J.; Zbasnik, J.

1986-08-01

This paper describes the construction and testing of the Mirror Fusion Test Facility superconducting magnet set. Construction of the first Yin Yang magnet was started in 1978. And although this particular magnet was later modified, the final construction of these magnets was not completed until 1985. When completed these 42 magnets weighed over 1200 tonnes and had a maximum stored energy of approximately 1200 MJ at full field. Together with power supplies, controls and liquid nitrogen radiation shields the cost of the fabrication of this system was over $100M. General Dynamics/Convair Division was responsible for the system design and the fabrication of 20 of the magnets. This contract was the largest single procurement action at the Lawrence Livermore National Laboratory. During the PACE acceptance tests, the 26 major magnets were operated at full field for more than 24 hours while other MFTF subsystems were tested. From all of the data, the magnets operated to the performance specifications. For physics operation in the future, additional helium and nitrogen leak checking and repair will be necessary. In this report we will discuss the operation and testing of the MFTF Magnet System, the world's largest superconducting magnet set built to date. The topics covered include a schedule of the major events, summary of the fabrication work, summary of the installation work, summary of testing and test results, and lessons learned

Applications of Research Reactors Towards Research on Materials for Nuclear Fusion Technology. Proceedings of a Technical Meeting

International Nuclear Information System (INIS)

2013-11-01

Controlled nuclear fusion is widely considered to represent a nearly unlimited source of energy. Recent progress in the quest for fusion energy includes the design and current construction of the International Thermonuclear Experimental Reactor (ITER), for which a licence has recently been obtained as a first of its kind fusion nuclear installation. ITER is designed to demonstrate the scientific and technological feasibility of fusion energy production in excess of 500 MW for several consecutive minutes. ITER, however, will not be able to address all the nuclear fusion technology issues associated with the design, construction and operation of a commercial fusion power plant. The demonstration of an adequate tritium or fuel breeding ratio, as well as the development, characterization and testing of structural and functional materials in an integrated nuclear fusion environment, are examples of issues for which ITER is unable to deliver complete answers. To fill this knowledge gap, several facilities are being discussed, such as the International Fusion Materials Irradiation Facility and, eventually, a fusion demonstration power plant (DEMO). However, for these facilities, a vast body of preliminary research remains to be performed, for instance, concerning the preselection and testing of suitable materials able to withstand the high temperature and pressure, and intense radiation environment of a fusion reactor. Given their capacity for material testing in terms of available intense neutron fluxes, dedicated irradiation facilities and post-irradiation examination laboratories, high flux research reactors or material test reactors (MTRs) will play an indispensable role in the development of fusion technology. Moreover, research reactors have already achieved an esteemed legacy in the understanding of material properties and behaviour, and the knowledge gained from experiments in fission materials in certain cases can be applied to fusion systems, particularly those

Fusion energy for the 21st century

International Nuclear Information System (INIS)

Harris, J.H.

1999-01-01

Fusion reactions like those that power the stars have the potential of providing bulk electricity generation with reduced emissions and low radioactive hazard, but pose many challenges in physics and technology. The H-1 Heliac Major National Research Facility now being developed offers Australian scientists and engineers an opportunity to participate in the collaborative international fusion research program. Work on H-1NF contributes not only to the realisation of fusion power, but offers the stimulus and opportunity for advanced training and the development of spin-off technology

Comparison of material irradiation conditions for fusion, spallation, stripping and fission neutron sources

International Nuclear Information System (INIS)

Vladimirov, P.; Moeslang, A.

2004-01-01

Selection and development of materials capable of sustaining irradiation conditions expected for a future fusion power reactor remain a big challenge for material scientists. Design of other nuclear facilities either in support of the fusion materials testing program or for other scientific purposes presents a similar problem of irradiation resistant material development. The present study is devoted to an evaluation of the irradiation conditions for IFMIF, ESS, XADS, DEMO and typical fission reactors to provide a basis for comparison of the data obtained for different material investigation programs. The results obtained confirm that no facility, except IFMIF, could fit all user requirements imposed for a facility for simulation of the fusion irradiation conditions

Inertial fusion in the nineties

International Nuclear Information System (INIS)

Harris, D.; Dudziak, D.J.; Cartwright, D.C.

1987-01-01

The 1980s have proven to be an exciting time for the inertial confinement fusion (ICF) program. Major new laser and light-ion drivers have been constructed and have produced some encouraging results. The 1990s will be a crucial time for the ICF program. A decision for proceeding with the next facility is scheduled for the early 1990s. If the decision is positive, planning and construction of this facility will occur. Depending on the time required for design and construction, this next-generation facility could become operational near the turn of the century

Tritium and ignition target management at the National Ignition Facility.

Science.gov (United States)

Draggoo, Vaughn

2013-06-01

Isotopic mixtures of hydrogen constitute the basic fuel for fusion targets of the National Ignition Facility (NIF). A typical NIF fusion target shot requires approximately 0.5 mmoles of hydrogen gas and as much as 750 GBq (20 Ci) of 3H. Isotopic mix ratios are specified according to the experimental shot/test plan and the associated test objectives. The hydrogen isotopic concentrations, absolute amounts, gas purity, configuration of the target, and the physical configuration of the NIF facility are all parameters and conditions that must be managed to ensure the quality and safety of operations. An essential and key step in the preparation of an ignition target is the formation of a ~60 μm thick hydrogen "ice" layer on the inner surface of the target capsule. The Cryogenic Target Positioning System (Cryo-Tarpos) provides gas handling, cyro-cooling, x-ray imaging systems, and related instrumentation to control the volumes and temperatures of the multiphase (solid, liquid, and gas) hydrogen as the gas is condensed to liquid, admitted to the capsule, and frozen as a single spherical crystal of hydrogen in the capsule. The hydrogen fuel gas is prepared in discrete 1.7 cc aliquots in the LLNL Tritium Facility for each ignition shot. Post-shot hydrogen gas is recovered in the NIF Tritium Processing System (TPS). Gas handling systems, instrumentation and analytic equipment, material accounting information systems, and the shot planning systems must work together to ensure that operational and safety requirements are met.

Inertial Fusion Driven By Intense Heavy-Ion Beams

International Nuclear Information System (INIS)

Sharp, W.M.; Friedman, A.; Grote, D.P.; Barnard, J.J.; Cohen, R.H.; Dorf, M.A.; Lund, S.M.; Perkins, L.J.; Terry, M.R.; Logan, B.G.; Bieniosek, F.M.; Faltens, A.; Henestroza, E.; Jung, J.Y.; Kwan, J.W.; Lee, E.P.; Lidia, S.M.; Ni, P.A.; Reginato, L.L.; Roy, P.K.; Seidl, P.A.; Takakuwa, J.H.; Vay, J.-L.; Waldron, W.L.; Davidson, R.C.; Gilson, E.P.; Kaganovich, I.D.; Qin, H.; Startsev, E.; Haber, I.; Kishek, R.A.; Koniges, A.E.

2011-01-01

Intense heavy-ion beams have long been considered a promising driver option for inertial-fusion energy production. This paper briefly compares inertial confinement fusion (ICF) to the more-familiar magnetic-confinement approach and presents some advantages of using beams of heavy ions to drive ICF instead of lasers. Key design choices in heavy-ion fusion (HIF) facilities are discussed, particularly the type of accelerator. We then review experiments carried out at Lawrence Berkeley National Laboratory (LBNL) over the past thirty years to understand various aspects of HIF driver physics. A brief review follows of present HIF research in the US and abroad, focusing on a new facility, NDCX-II, being built at LBNL to study the physics of warm dense matter heated by ions, as well as aspects of HIF target physics. Future research directions are briefly summarized.

Safety assessment for the rf Test Facility

International Nuclear Information System (INIS)

Nagy, A.; Beane, F.

1984-08-01

The Radio Frequency Test Facility (RFTF) is a part of the Magnetic Fusion Program's rf Heating Experiments. The goal of the Magnetic Fusion Program (MFP) is to develop and demonstrate the practical application of fusion. RFTF is an experimental device which will provide an essential link in the research effort aiming at the realization of fusion power. This report was compiled as a summary of the analysis done to ensure the safe operation of RFTF

Developmental potential of bovine hand-made clone embryos reconstructed by aggregation or fusion with distinct cytoplasmic volumes.

Science.gov (United States)

Ribeiro, Eduardo de Souza; Gerger, Renato Pereira da Costa; Ohlweiler, Lain Uriel; Ortigari, Ivens; Mezzalira, Joana Cláudia; Forell, Fabiana; Bertolini, Luciana Relly; Rodrigues, José Luiz; Ambrósio, Carlos Eduardo; Miglino, Maria Angélica; Mezzalira, Alceu; Bertolini, Marcelo

2009-09-01

Animal cloning has been associated with developmental abnormalities, with the level of heteroplasmy caused by the procedure being one of its potential limiting factors. The aim of this study was to determine the effect of the fusion of hemicytoplasts or aggregation of hemiembryos, varying the final cytoplasmic volume, on development and cell density of embryos produced by hand-made cloning (HMC), parthenogenesis or by in vitro fertilization (IVF). One or two enucleated hemicytoplasts were paired and fused with one skin somatic cell. Activated clone and zona-free parthenote embryos and hemiembryos were in vitro cultured in the well-of-the-well (WOW) system, being allocated to one of six experimental groups, on a per WOW basis: single clone or parthenote hemiembryos (1 x 50%); aggregation of two (2 x 50%), three (3 x 50%), or four (4 x 50%) clone or parthenote hemiembryos; single clone or parthenote embryos (1 x 100%); or aggregation of two clone or parthenote embryos (2 x 100%). Control zona-intact parthenote or IVF embryos were in vitro cultured in four-well dishes. Results indicated that the increase in the number of aggregated structures within each WOW was followed by a linear increase in cleavage, blastocyst rate, and cell density. The increase in cytoplasmic volume, either by fusion or by aggregation, had a positive effect on embryo development, supporting the establishment of pregnancies and the birth of a viable clone calf after transfer to recipients. However, embryo aggregation did not improve development on a hemicytoplast basis, except for the aggregation of two clone embryos.

Fusion reactor design and technology 1986. V. 1

International Nuclear Information System (INIS)

1987-01-01

The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

Performance characterization of the FLEX low pressure helium facility for fusion technology experiments

Energy Technology Data Exchange (ETDEWEB)

Schlindwein, Georg, E-mail: schlindwein@kit.edu; Arbeiter, Frederik

2014-10-15

Highlights: • A gas loop for fusion R and D has been built and tested. • Facility requirements and their implementation are given. • The loop's functions and instrumentation are explained. • The loops performance has been characterized. - Abstract: FLEX (Fluid Dynamics Experimental Facility) is a multi-purpose small scale gas loop for research on fluid and thermodynamic investigations, especially heat transfer, flow field measurements and gas purification. Initially it was built for investigation on mini-channel gas-flow to design the HFTM module of IFMIF. Because of its versatility it offers a wide range of further applications, e.g. the research of pressure drops in mockups of breeder units of the helium cooled pebble bed (HCPB) test blanket module for ITER. The main parameters of the loop, which can be operated with inert gases and air are: (i) operation gas pressure 0.02–0.38 MPa abs., (ii) test section pressure head up to 0.12 MPa, (iii) tolerable gas temperature RT – 200 °C and (iv) mass flow rate 0.2–12 × 10{sup −3} kg/s for Helium. This paper gives a detailed view of the loop assembly with the components that generate and regulate the mass flow and loop pressure. The measurement instrumentation will be presented as well as a representative mass flow-pressure drop characteristic. Furthermore, the achievable gas purity will be discussed.

The INEL Tritium Research Facility

International Nuclear Information System (INIS)

Longhurst, G.R.

1990-01-01

The Tritium Research Facility (TRF) at the Idaho National Engineering Laboratory (INEL) is a small, multi-user facility dedicated to research into processes and phenomena associated with interaction of hydrogen isotopes with other materials. Focusing on bench-scale experiments, the main objectives include resolution of issues related to tritium safety in fusion reactors and the science and technology pertinent to some of those issues. In this report the TRF and many of its capabilities will be described. Work presently or recently underway there will be discussed, and the implications of that work to the development of fusion energy systems will be considered. (orig.)

The INEL Tritium Research Facility

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R. (Idaho National Engineering Lab., Idaho Falls (USA))

1990-06-01

The Tritium Research Facility (TRF) at the Idaho National Engineering Laboratory (INEL) is a small, multi-user facility dedicated to research into processes and phenomena associated with interaction of hydrogen isotopes with other materials. Focusing on bench-scale experiments, the main objectives include resolution of issues related to tritium safety in fusion reactors and the science and technology pertinent to some of those issues. In this report the TRF and many of its capabilities will be described. Work presently or recently underway there will be discussed, and the implications of that work to the development of fusion energy systems will be considered. (orig.).

Conditioning and handling of tritiated wastes at Canadian nuclear power facilities

International Nuclear Information System (INIS)

Krochmalnek, L.S.; Krasznai, J.P.; Carney, M.

1987-04-01

Ontario Hydro operates a 10,000 MW capacity nuclear power system utilizing the CANDU pressurized heavy water reactor design. The use of D 2 O as moderator and coolant results in the production of about 2400 Ci of tritium per MWe-yr. As a result, there is significant Canadian experience in the treatment, handling, transport and storage of tritiated wastes. Ontario Hydro operates its own reactor waste storage site which includes systems for volume reduction, immobilization and packaging of wastes. In addition, a facility to remove tritium from heavy water is presently being commissioned at the Darlington nuclear site. This facility will generate tritiated liquid and solid waste that will have to be properly conditioned prior to storage or disposal. The nature of these various wastes and the processes/packaging required to meet storage/disposal criteria are judged to have relevance to investigations in fusion facility waste arisings. Experience to date, planned operational procedures and ongoing R and D in this area are described

Fusion blanket testing in MFTF-α + T

International Nuclear Information System (INIS)

Kleefeldt, K.

1985-01-01

The Mirror Fusion Test Facility-α + T (MFTF-α + T) is an upgraded version of the current MFTF-B test facility at Lawrence Livermore National Laboratory, and is designed for near-term fusion-technology-integrated tests at a neutron flux of 2 MW/m 2 . Currently, the fusion community is screening blanket and related issues to determine which ones can be addressed using MFTF-α + T. In this work, the minimum testing needs to address these issues are identified for the liquid-metal-cooled blanket and the solid-breeder blanket. Based on the testing needs and on the MFTF-α + T capability, a test plan is proposed for three options; each option covers a six to seven year testing phase. The options reflect the unresolved question of whether to place the research and development (R and D) emphasis on liquid-metal or solid-breeder blankets. In each case, most of the issues discussed can be addressed to a reasonable extent in MFTF-α+T

Development of fusion fuel cycles: Large deviations from US defense program systems

Energy Technology Data Exchange (ETDEWEB)

Klein, James Edward, E-mail: james.klein@srnl.doe.gov; Poore, Anita Sue; Babineau, David W.

2015-10-15

Highlights: • All tritium fuel cycles start with a “Tritium Process.” All have similar tritium processing steps. • Fusion tritium fuel cycles minimize process tritium inventories for various reasons. • US defense program facility designs did not minimize in-process inventories. • Reduced inventory tritium facilities will lower public risk. - Abstract: Fusion energy research is dominated by plasma physics and materials technology development needs with smaller levels of effort and funding dedicated to tritium fuel cycle development. The fuel cycle is necessary to supply and recycle tritium at the required throughput rate; additionally, tritium confinement throughout the facility is needed to meet regulatory and environmental release limits. Small fuel cycle development efforts are sometimes rationalized by stating that tritium processing technology has already been developed by nuclear weapons programs and these existing processes only need rescaling or engineering design to meet the needs of fusion fuel cycles. This paper compares and contrasts features of tritium fusion fuel cycles to United States Cold War era defense program tritium systems. It is concluded that further tritium fuel cycle development activities are needed to provide technology development beneficial to both fusion and defense programs tritium systems.

The National Ignition Facility and Industry

Science.gov (United States)

Harri, J. G.; Paisner, J. A.; Lowdermilk, W. H.; Boyes, J. D.; Kumpan, S. A.; Sorem, M. S.

1994-09-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. The National Ignition Facility construction project will require the best of our construction industries and its success will depend on the best products offered by hundreds of the nation's high technology companies. Three-fourths of the construction costs will be invested in industry. This article reviews the design, cost and schedule, and required industrial involvement associated with the construction project.

The National Ignition Facility and industry

International Nuclear Information System (INIS)

Harri, J.G.; Lowdermilk, W.H.; Paisner, J.A.; Boyes, J.D.; Kumpan, S.A.; Sorem, M.S.

1994-01-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. The National Ignition Facility construction project will require the best of national construction industries and its success will depend on the best products offered by hundreds of the nation's high technology companies. Three-fourths of the construction costs will be invested in industry. This article reviews the design, cost and schedule, and required industrial involvement associated with the construction project

Report of the Fusion Energy Sciences Advisory Committee. Panel on Integrated Simulation and Optimization of Magnetic Fusion Systems