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

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

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

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

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

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

Intense neutron irradiation facility for fusion reactor materials

Energy Technology Data Exchange (ETDEWEB)

Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

Demountable toroidal fusion core facility for physics optimization and fusion engineering

International Nuclear Information System (INIS)

Bogart, S.L.; Wagner, C.E.; Krall, N.A.; Dalessandro, J.A.; Weggel, C.F.; Lund, K.O.; Sedehi, S.

1986-01-01

Following a successful compact ignition tokamak (CIT) experiment, a fusion facility will be required for physics optimization (POF) and fusion engineering research (FERF). The POF will address issues such as high-beta operation, current drive, impurity control, and will test geometric and configurational variations such as the spherical torus or the reversed-field pinch (RFP). The FERF will be designed to accumulate rapidly a large neutron dose in prototypical fusion subsystems exposed to radiation. Both facilities will require low-cost replacement cores and rapid replacement times. The Demountable Toroidal Fusion Core (DTFC) facility is designed to fulfill these requirements. It would be a cost-effective stepping stone between the CIT and a demonstration fusion reactor

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

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 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

Near term, low cost, 14 MeV fusion neutron irradiation facility for testing the viability of fusion structural materials

Energy Technology Data Exchange (ETDEWEB)

Kulcinski, Gerald L., E-mail: glkulcin@wisc.edu [University of Wisconsin-Madison, Madison, WI (United States); Radel, Ross F. [Phoenix Nuclear Labs LLC, Monona, WI (United States); Davis, Andrew [University of Wisconsin-Madison, Madison, WI (United States)

2016-11-01

For over 50 years, engineers have been looking for an irradiation facility that can provide a fusion reactor appropriate neutron spectrum over a significant volume to test fusion reactor materials that is relatively inexpensive and can be built in a minimum of time. The 14 MeV neutron irradiation facility described here can nearly exactly duplicate the neutron spectrum typical of a DT fusion reactor first wall at damage rates of ≈4 displacements per atom and 40 appm He generated over a 2 l volume per full power year of operation. The projected cost of this multi-beam facility is estimated at ≈$20 million and it can be built in <4 years. A single-beam prototype, funded by the U.S. Department of Energy, is already being built to produce medical isotopes. The neutrons are produced by a 300 keV deuterium beam accelerated into 4 kPa (30 Torr) tritium target. The total tritium inventory is <2 g and <0.1 g of T{sub 2} is consumed per year. The core technology proposed has already been fully demonstrated, and no new plasma physics or materials innovations will be required for the test facility to become operational.

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

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

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

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.

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

An Overview of INEL Fusion Safety R&D Facilities

Science.gov (United States)

McCarthy, K. A.; Smolik, G. R.; Anderl, R. A.; Carmack, W. J.; Longhurst, G. R.

1997-06-01

The Fusion Safety Program at the Idaho National Engineering Laboratory has the lead for fusion safety work in the United States. Over the years, we have developed several experimental facilities to provide data for fusion reactor safety analyses. We now have four major experimental facilities that provide data for use in safety assessments. The Steam-Reactivity Measurement System measures hydrogen generation rates and tritium mobilization rates in high-temperature (up to 1200°C) fusion relevant materials exposed to steam. The Volatilization of Activation Product Oxides Reactor Facility provides information on mobilization and transport and chemical reactivity of fusion relevant materials at high temperature (up to 1200°C) in an oxidizing environment (air or steam). The Fusion Aerosol Source Test Facility is a scaled-up version of VAPOR. The ion-implanta-tion/thermal-desorption system is dedicated to research into processes and phenomena associated with the interaction of hydrogen isotopes with fusion materials. In this paper we describe the capabilities of these facilities.

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.

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

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

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.)

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)

Safety of magnetic fusion facilities: Guidance

International Nuclear Information System (INIS)

1996-05-01

This document provides guidance for the implementation of the requirements identified in DOE-STD-6002-96, Safety of Magnetic Fusion Facilities: Requirements. This guidance is intended for the managers, designers, operators, and other personnel with safety responsibilities for facilities designated as magnetic fusion facilities. While the requirements in DOE-STD-6002-96 are generally applicable to a wide range of fusion facilities, this Standard, DOE-STD-6003-96, is concerned mainly with the implementation of those requirements in large facilities such as the International Thermonuclear Experimental Reactor (ITER). Using a risk-based prioritization, the concepts presented here may also be applied to other magnetic fusion facilities. This Standard is oriented toward regulation in the Department of Energy (DOE) environment as opposed to regulation by other regulatory agencies. As the need for guidance involving other types of fusion facilities or other regulatory environments emerges, additional guidance volumes should be prepared. The concepts, processes, and recommendations set forth here are for guidance only. They will contribute to safety at magnetic fusion facilities

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

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

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

Test facility TIMO for testing the ITER model cryopump

International Nuclear Information System (INIS)

Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

2001-01-01

Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L' Air Liquide in the form of a NET contract. (author)

Test facility TIMO for testing the ITER model cryopump

International Nuclear Information System (INIS)

Haas, H.; Day, C.; Mack, A.; Methe, S.; Boissin, J.C.; Schummer, P.; Murdoch, D.K.

1999-01-01

Within the framework of the European Fusion Technology Programme, FZK is involved in the research and development process for a vacuum pump system of a future fusion reactor. As a result of these activities, the concept and the necessary requirements for the primary vacuum system of the ITER fusion reactor were defined. Continuing that development process, FZK has been preparing the test facility TIMO (Test facility for ITER Model pump) since 1996. This test facility provides for testing a cryopump all needed infrastructure as for example a process gas supply including a metering system, a test vessel, the cryogenic supply for the different temperature levels and a gas analysing system. For manufacturing the ITER model pump an order was given to the company L'Air Liquide in the form of a NET contract. (author)

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

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

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

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.

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

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)

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.

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.)

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

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

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

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.)

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

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

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

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.

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

HFR irradiation testing of fusion materials

International Nuclear Information System (INIS)

Conrad, R.; von der Hardt, P.; Loelgen, R.; Scheurer, H.; Zeisser, P.

1984-01-01

The present and future role of the High Flux Reactor Petten for fusion materials testing has been assessed. For practical purposes the Tokamak-based fusion reactor is chosen as a point of departure to identify material problems and materials data needs. The identification is largely based on the INTOR and NET design studies, the reported programme strategies of Japan, the U.S.A. and the European Communities for technical development of thermonuclear fusion reactors and on interviews with several experts. Existing and planned irradiation facilities, their capabilities and limitations concerning materials testing have been surveyed and discussed. It is concluded that fission reactors can supply important contributions for fusion materials testing. From the point of view of future availability of fission testing reactors and their performance it appears that the HFR is a useful tool for materials testing for a large variety of materials. Prospects and recommendations for future developments are given

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

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

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

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

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

Safety of magnetic fusion facilities: Volume 2, Guidance

International Nuclear Information System (INIS)

1995-01-01

This document provides guidance for the implementation of the requirements identified in Vol. 1 of this Standard. This guidance is intended for the managers, designers, operators, and other personnel with safety responsibilities for facilities designated as magnetic fusion facilities. While Vol. 1 is generally applicable in that requirements there apply to a wide range of fusion facilities, this volume is concerned mainly with large facilities such as the International Thermonuclear Experimental Reactor (ITER). Using a risk-based prioritization, the concepts presented here may also be applied to other magnetic fusion facilities. This volume is oriented toward regulation in the Department of Energy (DOE) environment

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

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

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

International Nuclear Information System (INIS)

Abdou, M.

1984-10-01

This volume contains the following chapters (1) neutronics tests, (2) fluence considerations, (3) instrumentation and test matrix, (4) non-neutron test stands, (5) accelerator-based point neutron sources, (6) utilization of fission reactors, (7) tandem mirror test facilities, (8) tokamak fusion test facilities, (9) reliability development testing impacts on fusion reactor availability, and (10) fusion development scenarios. In addition, the following appendices are included: (1) evaluation of experience from fast breeder reactors, (2) observations of experts from the fission field, (3) evaluation of experience from the aerospace industry, (4) characterization of fusion nuclear systems operating environment, (5) modelling of MFTF-α+T high gamma mode performance, and (6) small-scale, multiple effects testing at US/DOE breeder reactor in-pile facilities

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)

IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

International Nuclear Information System (INIS)

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

2002-03-01

Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D + )-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m 2 , 20 dpa/year for Fe) in a volume of 500 cm 3 for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (author)

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

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

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

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

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)

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

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

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

IFMIF (International Fusion Materials Irradiation Facility) key element technology phase interim report

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hiroo; Ida, Mizuho; Sugimoto, Masayoshi; Takeuchi, Hiroshi; Yutani, Toshiaki (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-03-01

Activities of International Fusion Materials Irradiation Facility (IFMIF) have been performed under an IEA collaboration since 1995. IFMIF is an accelerator-based deuteron (D{sup +})-lithium (Li) neutron source designed to produce an intense neutron field (2 MW/m{sup 2}, 20 dpa/year for Fe) in a volume of 500 cm{sup 3} for testing candidate fusion materials. In 2000, a 3 year Key Element technology Phase (KEP) of IFMIF was started to reduce the key technology risk factors. This interim report summarizes the KEP activities until mid 2001 in the major project work-breakdown areas of accelerator, target, test facilities and design integration. (author)

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

LLNL superconducting magnets test facility

Energy Technology Data Exchange (ETDEWEB)

Manahan, R; Martovetsky, N; Moller, J; Zbasnik, J

1999-09-16

The FENIX facility at Lawrence Livermore National Laboratory was upgraded and refurbished in 1996-1998 for testing CICC superconducting magnets. The FENIX facility was used for superconducting high current, short sample tests for fusion programs in the late 1980s--early 1990s. The new facility includes a 4-m diameter vacuum vessel, two refrigerators, a 40 kA, 42 V computer controlled power supply, a new switchyard with a dump resistor, a new helium distribution valve box, several sets of power leads, data acquisition system and other auxiliary systems, which provide a lot of flexibility in testing of a wide variety of superconducting magnets in a wide range of parameters. The detailed parameters and capabilities of this test facility and its systems are described in the paper.

R and D needs assessment for the Engineering Test Facility

International Nuclear Information System (INIS)

1980-10-01

The Engineering Test Facility (ETF), planned to be the next major US magnetic fusion device, has its mission (1) to provide the capability for moving into the engineering phase of fusion development and (2) to provide a test-bed for reactor components in a fusion environment. The design, construction, and operation of the ETF requires an increasing emphasis on certain key research and development (R and D) programs in magnetic fusion in order to provide the necessary facility design base. This report identifies these needs and discusses the apparent inadequacies of the presently planned US program to meet them, commensurate with the ETF schedule

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

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

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...

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

ORNL 150 keV neutral beam test facility

International Nuclear Information System (INIS)

Gardner, W.L.; Kim, J.; Menon, M.M.; Schilling, G.

1977-01-01

The 150 keV neutral beam test facility provides for the testing and development of neutral beam injectors and beam systems of the class that will be needed for the Tokamak Fusion Test Reactor (TFTR) and The Next Step (TNS). The test facility can simulate a complete beam line injection system and can provide a wide range of experimental operating conditions. Herein is offered a general description of the facility's capabilities and a discussion of present system performance

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

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

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.)

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

Decontamination and decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Walton, G.R.; Perry, E.D.; Commander, J.C.; Spampinato, P.T.

1994-01-01

The Tokamak Fusion Test Reactor (TFTR) is scheduled to complete its end-of-life deuterium-tritium (D-T) experiments in September 1994. The D-T operation will result in the TFTR machine structure becoming activated, and plasma facing and vacuum components will be contaminated with tritium. The resulting machine activation levels after a two year cooldown period will allow hands on dismantling for external structures, but require remote dismantling for the vacuum vessel. The primary objective of the Decontamination and Decommissioning (D ampersand D) Project is to provide a facility for construction of a new Department of Energy (DOE) experimental fusion reactor by March 1998. The project schedule calls for a two year shutdown period when tritium decontamination of the vacuum vessel, neutral beam injectors and other components will occur. Shutdown will be followed by an 18 month period of D ampersand D operations. The technical objectives of the project are to: safely dismantle and remove components from the test cell complex; package disassembled components in accordance with applicable regulations; ship packages to a DOE approved disposal or material recycling site; and develop expertise using remote disassembly techniques on a large scale fusion facility. This paper discusses the D ampersand D objectives, the facility to be decommissioned, and the technical plan that will be implemented

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

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)

Stability tests of the Westinghouse coil in the International Fusion Superconducting Magnet Test Facility

International Nuclear Information System (INIS)

Dresner, L.; Fehling, D.T.; Lubell, M.S.; Lue, J.W.; Luton, J.N.; McManamy, T.J.; Shen, S.S.; Wilson, C.T.

1987-09-01

The Westinghouse coil is one of three forced-flow coils in the six-coil toroidal array of the International Fusion Superconducting Magnet Test Facility at Oak Ridge National Laboratory. It is wound with an 18-kA, Nb 3 Sn/Cu, cable-in-conduit superconductor structurally supported by aluminum plates and cooled by 4-K, 15-atm supercritical helium. The coil is instrumented to permit measurement of helium temperature, pressure, and flow rate; structure temperature and strain; field; and normal zone voltage. A resistive heater has been installed to simulate nuclear heating, and inductive heaters have been installed to facilitate stability testing. The coil has been tested both individually and in the six-coil array. The tests covered charging to full design current and field, measuring the current-sharing threshold temperature using the resistive heaters, and measuring the stability margin using the pulsed inductive heaters. At least one section of the conductor exhibits a very broad resistive transition (resistive transition index = 4). The broad transition, though causing the appearance of voltage at relatively low temperatures, does not compromise the stability margin of the coil, which was greater than 1.1 J/cm 3 of strands. In another, nonresistive location, the stability margin was between 1.7 and 1.9 J/cm 3 of strands. The coil is completely stable in operation at 100% design current in both the single- and six-coil modes

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.)

IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

International Nuclear Information System (INIS)

2003-03-01

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m 2 , 20 dpa/y in Fe, in a volume of 500 cm 3 and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

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.

Development of an engineering-scale nuclear test of a solid-breeder fusion-blanket concept

International Nuclear Information System (INIS)

Deis, G.A.; Bohn, T.S.; Hsu, P.Y.; Miller, L.G.; Scott, A.J.; Watts, K.D.; Welch, E.C.

1983-08-01

As part of the Phase I effort on Program Element-II (PE-II) of the Office of Fusion Energy/Argonne National Laboratory First Wall/Blanket/Shield Engineering Technology Program, a study has been performed to develop preconceptual hardware designs and preliminary test program descriptions for two fission-reactor-based tests of a water-cooled, solid-breeder fusion reactor blanket concept. First, a list of potentially acceptable reactor facilities is developed, based on a list of required reactor characteristics. From this set of facilities, two facilities are selected for study: the Oak Ridge Research Reactor (ORR) and the Power Burst Facility (PBF). A test which employs a cylindrical unit cell of a solid-breeder fusion reactor blanket, with pressurized-water cooling is designed for each facility. The test design is adjusted to the particular characteristics of each reactor. These two test designs are then compared on the basis of technical issues and cost. Both tests can satisfy the PE-II mission: blanket thermal hydraulic and thermomechanical issues. In addition, both reactors will produce prototypical tritium production rates and profiles and release characteristics with little or no additional modifications

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

Background field coils for the High Field Test Facility

International Nuclear Information System (INIS)

Zbasnik, J.P.; Cornish, D.N.; Scanlan, R.M.; Jewell, A.M.; Leber, R.L.; Rosdahl, A.R.; Chaplin, M.R.

1980-01-01

The High Field Test Facility (HFTF), presently under construction at LLNL, is a set of superconducting coils that will be used to test 1-m-o.d. coils of prototype conductors for fusion magnets in fields up to 12 T. The facility consists of two concentric sets of coils; the outer set is a stack of Nb-Ti solenoids, and the inner set is a pair of solenoids made of cryogenically-stabilized, multifilamentary Nb 3 Sn superconductor, developed for use in mirror-fusion magnets. The HFTF system is designed to be parted along the midplane to allow high-field conductors, under development for Tokamak fusion machines, to be inserted and tested. The background field coils were wound pancake-fashion, with cold-welded joints at both the inner and outer diameters. Turn-to-turn insulation was fabricated at LLNL from epoxy-fiberglass strip. The coils were assembled and tested in our 2-m-diam cryostat to verify their operation

IFMIF-KEP. International fusion materials irradiation facility key element technology phase report

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-03-01

The International Fusion Materials Irradiation Facility (IFMIF) is an accelerator-based D-Li neutron source designed to produce an intense neutron field that will simulate the neutron environment of a D-T fusion reactor. IFMIF will provide a neutron flux equivalent to 2 MW/m{sup 2}, 20 dpa/y in Fe, in a volume of 500 cm{sup 3} and will be used in the development and qualification of materials for fusion systems. The design activities of IFMIF are performed under an IEA collaboration which began in 1995. In 2000, a three-year Key Element Technology Phase (KEP) of IFMIF was undertaken to reduce the key technology risk factors. This KEP report describes the results of the three-year KEP activities in the major project areas of accelerator, target, test facilities and design integration. (author)

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

Material testing facilities and programs for plasma-facing component testing

Science.gov (United States)

Linsmeier, Ch.; Unterberg, B.; Coenen, J. W.; Doerner, R. P.; Greuner, H.; Kreter, A.; Linke, J.; Maier, H.

2017-09-01

Component development for operation in a large-scale fusion device requires thorough testing and qualification for the intended operational conditions. In particular environments are necessary which are comparable to the real operation conditions, allowing at the same time for in situ/in vacuo diagnostics and flexible operation, even beyond design limits during the testing. Various electron and neutral particle devices provide the capabilities for high heat load tests, suited for material samples and components from lab-scale dimensions up to full-size parts, containing toxic materials like beryllium, and being activated by neutron irradiation. To simulate the conditions specific to a fusion plasma both at the first wall and in the divertor of fusion devices, linear plasma devices allow for a test of erosion and hydrogen isotope recycling behavior under well-defined and controlled conditions. Finally, the complex conditions in a fusion device (including the effects caused by magnetic fields) are exploited for component and material tests by exposing test mock-ups or material samples to a fusion plasma by manipulator systems. They allow for easy exchange of test pieces in a tokamak or stellarator device, without opening the vessel. Such a chain of test devices and qualification procedures is required for the development of plasma-facing components which then can be successfully operated in future fusion power devices. The various available as well as newly planned devices and test stands, together with their specific capabilities, are presented in this manuscript. Results from experimental programs on test facilities illustrate their significance for the qualification of plasma-facing materials and components. An extended set of references provides access to the current status of material and component testing capabilities in the international fusion programs.

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)

Current Status and Performance Tests of Korea Heat Load Test Facility KoHLT-EB

Energy Technology Data Exchange (ETDEWEB)

Kim, Sukkwon; Jin, Hyunggon; Shin, Kyuin; Choi, Boguen; Lee, Eohwak; Yoon, Jaesung; Lee, Dongwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Duckhoi; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2013-05-15

A commissioning test has been scheduled to establish the installation and preliminary performance experiments of the copper hypervapotron mockups. And a qualification test will be performed to evaluate the CuCrZr duct liner in the ITER neutral beam injection facility and the ITER first wall small-scale mockups of the semi-prototype, at up to 1.5 and 5 MW/m{sup 2} high heat flux. Also, this system will be used to test other PFCs for ITER and materials for tokamak reactors. Korean high heat flux test facility(KoHLT-EB; Korea Heat Load Test facility - Electron Beam) by using an electron beam system has been constructed in KAERI to perform the qualification test for ITER blanket FW semi-prototype mockups, hypervapotron cooling devices in fusion devices, and other ITER plasma facing components. The commissioning and performance tests with the supplier of e-gun system have been performed on November 2012. The high heat flux test for hypervapotron cooling device and calorimetry were performed to measure the surface heat flux, the temperature profile and cooling performance. Korean high heat flux test facility for the plasma facing components of nuclear fusion machines will be constructed to evaluate the performance of each component. This facility for the plasma facing materials will be equipped with an electron beam system with a 60 kV acceleration gun.

Magnets for the Mirror Fusion Test Facility: testing of the first Yin-Yang and the design and development of other magnets

International Nuclear Information System (INIS)

Kozman, T.A.; Wang, S.T.; Chang, Y.

1983-01-01

Completed in May 1981, the first Yin-Yang magnet for the tandem Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory (LLNL) was successfully tested in February 1982 to its full design field (7.68 T) and current (5775 A). Since that time, the entire magnet array has been reconfigured - from the original A-cell to an axicell design. The MFTF-B magnet array now contains a total of 26 large superconducting coils: 2 sets of yin-yang pairs, 2 sets of transition magnets (each containing two coils), 2 sets of axicell magnets (each containing three coils), and 12 central-cell solenoids. This paper chronicles recent magnet history - from te testing of the initial yin-yang set, through the design of the axicell configuration, to the planned development of the system

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.)

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

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

Development and manufacture of a Nb3Sn superconductor for the high-field test facility

International Nuclear Information System (INIS)

Scanlan, R.M.; Cornish, D.N.; Spencer, C.R.; Gregory, E.; Adam, E.

1981-01-01

The High-Field Test Facility (HFTF) project has two primary goals. The first is to establish manufacturing capability for a Nb 3 Sn conductor suitable for use in a mirror fusion coil. The second is to provide a test facility for evaluating other fusion conductor designs at high fields. This paper describes some of the problems encountered and the solutions devised in working toward the first goal. Construction of the test facility coils will be described in a subsequent paper

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

A coil test facility for the cryogenic tests of the JT-60SA TF coils

International Nuclear Information System (INIS)

Chantant, M.; Genini, L.; Bayetti, P.; Millet, F.; Wanner, M.; Massaut, V.; Corte, A. Della; Ardelier-Desage, F.; Catherine-Dumont, V.; Dael, A.; Decool, P.; Donati, A.; Duchateau, J.L.; Garibaldi, P.; Girard, S.; Hatchressian, J.C.; Fejoz, P.; Jamotton, P.; Jourdheuil, L.; Juster, F.P.

2011-01-01

In the framework of the Broader Approach Activities, the EU will deliver to Japan the 18 superconducting coils, which constitute the JT-60SA Toroidal field magnet. These 18 coils, manufactured by France and Italy, will be cold tested before shipping to Japan. For this purpose, the European Joint Undertaking for ITER, the Development of Fusion Energy ('Fusion for Energy', F4E) and the European Voluntary Contributors are collaborating to design and set-up a coil test facility (CTF) and to perform the acceptance test of the 18 JT-60SA Toroidal Field (TF) coils. The test facility is designed to test one coil at a time at nominal current and cryogenic temperature. The test of the first coil of each manufacturer includes a quench triggered by increasing the temperature. The project is presently in the detailed design phase.

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

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

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

Role of fission-reactor-testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-01-01

Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

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

Program user's manual: cryogen system for the analysis for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

1979-04-01

The Mirror Fusion Test Facility being designed and constructed at the Lawrence Livermore Laboratory requires a liquid helium liquefaction, storage, distribution, and recovery system and a liquid nitrogen storage and distribution system. To provide a powerful analytical tool to aid in the design evolution of this system through hardware, a thermodynamic fluid flow model was developed. This model allows the Lawrence Livermore Laboratory to verify that the design meets desired goals and to play what if games during the design evolution. For example, what if the helium flow rate is changed in the magnet liquid helium flow loop; how does this affect the temperature, fluid quality, and pressure. This manual provides all the information required to run all or portions of this program as desired. In addition, the program is constructed in a modular fashion so changes or modifications can be made easily to keep up with the evolving design

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

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

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.

Some applications of fission-based testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

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

1981-10-01

The testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for fusion testing when they do become available. Fission testing is capable of filling many gaps in fusion reactor design information, and should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, work was performed on the irradiation of magnet insulators. This work is continuing with an improved test environment. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, a conceptual design is presented for a fission-based Integrated Test Facility (ITF), which can accommodate entire wall/blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

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

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

Ten years operating experience at the Fast Flux Test Facility: A decade of excellence

International Nuclear Information System (INIS)

Swaim, D.J.; Waldo, J.B.; Farabee, O.A.

1991-07-01

The Fast Flux Test Facility is a 400 MW(t) fast reactor cooled by three sodium loops. The Fast Flux Test Facility is managed by the Westinghouse Hanford Company for the US Department of Energy. The Fast Flux Test Facility was designed and constructed to provide irradiation testing of fuels and materials for the US Department of Energy Liquid Metal Reactor research program. Facility activities have increased to include fusion power materials testing, passive safety testing, isotope production, and international collaboration. 5 figs

Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

Energy Technology Data Exchange (ETDEWEB)

Ware, A.G.; Longhurst, G.R.

1981-12-01

This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available.

Test program element II blanket and shield thermal-hydraulic and thermomechanical testing, experimental facility survey

International Nuclear Information System (INIS)

Ware, A.G.; Longhurst, G.R.

1981-12-01

This report presents results of a survey conducted by EG and G Idaho to determine facilities available to conduct thermal-hydraulic and thermomechanical testing for the Department of Energy Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program. In response to EG and G queries, twelve organizations (in addition to EG and G and General Atomic) expressed interest in providing experimental facilities. A variety of methods of supplying heat is available

ITER primary cryopump test facility

International Nuclear Information System (INIS)

Petersohn, N.; Mack, A.; Boissin, J.C.; Murdoc, D.

1998-01-01

A cryopump as ITER primary vacuum pump is being developed at FZK under the European fusion technology programme. The ITER vacuum system comprises of 16 cryopumps operating in a cyclic mode which fulfills the vacuum requirements in all ITER operation modes. Prior to the construction of a prototype cryopump, the concept is tested on a reduced scale model pump. To test the model pump, the TIMO facility is being built at FZK in which the model pump operation under ITER environmental conditions, except for tritium exposure, neutron irradiation and magnetic fields, can be simulated. The TIMO facility mainly consists of a test vessel for ITER divertor duct simulation, a 600 W refrigerator system supplying helium in the 5 K stage and a 30 kW helium supply system for the 80 K stage. The model pump test programme will be performed with regard to the pumping performance and cryogenic operation of the pump. The results of the model pump testing will lead to the design of the full scale ITER cryopump. (orig.)

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

Initial testing of the tritium systems at the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Anderson, J.L.; Sissingh, R.A.P.; Gentile, C.A.; Rossmassler, R.L.; Walters, R.T.; Voorhees, D.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) at Princeton will start its D-T experiments in late 1993, introducing and operating the tokamak with tritium in order to begin the study of burning plasma physics in D-T. Trace tritium injection experiments, using small amounts of tritium will begin in the fall of 1993. In preparation for these experiments, a series of tests with low concentrations of tritium inn deuterium have been performed as an initial qualification of the tritium systems. These tests began in April 1993. This paper describes the initial testing of the equipment in the TFTR tritium facility

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.

Technical critique on radiation test facilities for the CTR surface and materials program

International Nuclear Information System (INIS)

Persiani, P.J.

1975-02-01

Major radiation test facilities will be necessary in the near-term (5 years) and long-term (greater than 10 years) future for the timely development and understanding of fusion confinement systems and of prototype fusion power reactors. The study includes the technical justifications and requirements for CTR Neutron and Plasma Radiation Test Facilities. The initial technical critique covers the feasibility and design problems: in upgrading the performance of the accelerator-rotating (solid TiT) target systems, and in transforming the accelerator-supersonic jet target concept into a radiation testing facility. A scoping assessment on the potential of a pulsed high-beta plasma device (dense plasma focus) is introduced to explore plasma concepts as near-term neutron and plasma radiation sources for the CTR Surface and Materials Program. (U.S.)

Plan for decommissioning the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Spampinato, P.T.; Walton, G.R.

1993-01-01

The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D ampersand D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D ampersand D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D ampersand D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D ampersand D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates

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

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.

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)

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)

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

Overview of the IFMIF test facility design in IFMIF/EVEDA phase

International Nuclear Information System (INIS)

Tian, Kuo; Abou-Sena, Ali; Arbeiter, Frederik; García, Ángela; Gouat, Philippe; Heidinger, Roland; Heinzel, Volker; Ibarra, Ángel; Leysen, Willem; Mas, Avelino; Mittwollen, Martin; Möslang, Anton; Theile, Jürgen; Yamamoto, Michiyoshi; Yokomine, Takehiko

2015-01-01

Highlights: • This paper summarizes the current design status of IFMIF EVEDA test facility. • The principle functions of the test facility and key components are described. • The brief specifications of the systems and key components are addressed. - Abstract: The test facility (TF) is one of the three major facilities of the International Fusion Material Irradiation Facility (IFMIF). Engineering designs of TF main systems and key components have been initiated and developed in the IFMIF EVEDA (Engineering Validation and Engineering Design Activities) phase since 2007. The related work covers the designs of a test cell which is the meeting point of the TF and accelerator facility and lithium facility, a series of test modules for experiments under different irradiation conditions, an access cell to accommodate remote handling systems, four test module handling cells for test module processing and assembling, and test facility ancillary systems for engineering support on energy, media, and control infrastructure. This paper summarizes the principle functions, brief specifications, and the current design status of the above mentioned IFMIF TF systems and key components.

Computer control and data acquisition system for the Mirror Fusion Test Facility Ion Cyclotron Resonant Heating System (ICRH)

International Nuclear Information System (INIS)

Cheshire, D.L.; Thomas, R.A.

1985-01-01

The Lawrence Livermore National Laboratory (LLNL) large Mirror Fusion Test Facility (MFTF-B) will employ an Ion Cyclotron Resonant Heating (ICRH) system for plasma startup. As the MFTF-B Industrial Participant, TRW has responsibility for the ICRH system, including development of the data acquisition and control system. During the MFTF-B Supervisory Control and Diagnostic System (SCDS). For subsystem development and checkout at TRW, and for verification and acceptance testing at LLNL, the system will be run from a stand-alone computer system designed to simulate the functions of SCDS. The ''SCDS Simulator'' was developed originally for the MFTF-B ECRH System; descriptions of the hardware and software are updated in this paper. The computer control and data acquisition functions implemented for ICRH are described, including development status, and test schedule at TRW and at LLNL. The application software is written for the SCDS Simulator, but it is programmed in PASCAL and designed to facilitate conversion for use on the SCDS computers

Superconductors for fusion magnets tested under pulsed field in SULTAN

International Nuclear Information System (INIS)

Bruzzone, P.; Bottura, L.; Katheder, H.; Blau, B.; Rohleder, I.; Vecsey, G.

1995-01-01

The SULTAN III test facility has been upgraded with a pair of pulsed field coils to carry out AC losses and stability experiments under full operating loads on large size, fusion conductors for ITER. A fast data aquisition system records the conductor behaviour under fast field transient. The commissioning results of the pulsed coils and instrumentation are critically discussed and the test capability of the set up is assessed. (orig.)

Test facility for the evaluation of microwave transmission components

International Nuclear Information System (INIS)

Fong, C.G.; Poole, B.R.

1985-01-01

A Low Power Test Facility (LPTF) was developed to evaluate the performance of Electron Cyclotron Resonance Heating (ECRH) microwave transmission components for the Mirror Fusion Test Facility (MFTF-B). The facility generates 26 to 60 GHz in modes of TE 01 , TE 02 , or TE 03 launched at power levels of 1/2 milliwatt. The propagation of the rf as it radiates from either transmitting or secondary reflecting microwave transmission components is recorded by a discriminating crystal detector mechanically manipulated at constant radius in spherical coordinates. The facility is used to test, calibrate, and verify the design of overmoded, circular waveguide components, quasi-optical reflecting elements before high power use. The test facility consists of microwave sources and metering components, such as VSWR, power and frequency meters, a rectangular TE 10 to circular TE 01 mode transducer, mode filter, circular TE 01 to 2.5 in. diameter overmoded waveguide with mode converters for combination of TE 01 to TE 03 modes. This assembly then connects to a circular waveguide launcher or the waveguide component under test

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.

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.

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.

Critical need for MFE: the Alcator DX advanced divertor test facility

Science.gov (United States)

Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

2013-10-01

Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

HELCZA-High heat flux test facility for testing ITER EU first wall components.

Czech Academy of Sciences Publication Activity Database

Prokůpek, J.; Samec, K.; Jílek, R.; Gavila, P.; Neufuss, S.; Entler, Slavomír

2017-01-01

Roč. 124, November (2017), s. 187-190 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : HELCZA * High heat flux * Electron beam testing * Test facility * Plasma facing components * First wall * Divertora Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 www.sciencedirect.com/science/article/pii/S0920379617302818

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

Simulation of the fusion materials irradiation test facility lithium and heat transport systems for abnormal events study

International Nuclear Information System (INIS)

Carlson, W.F.; Elyashar, N.N.

1981-01-01

A digital computer model of Fusion Materials Irradiation Test Facility's heat transport system has been developed. The model utilizes a set of coupled differential equations to simulate the dynamic behavior of the primary and secondary heat transport loop systems. The model has been used to investigate the stability of the proposed control schemes for lithium temperature and flow rate and for an extensive study of equipment failures and malfunction analysis. It was determined that certain equipment failures and malfunctions in the primary loop require a response from the control system within less than one second of the occurrence of the failure. The effects of equipment failures in the secondary loop were found to be less dramatic than the equivalent failures in the primary loop. The failures in the secondary loop generally required control action in time frames of the order of minutes

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

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.

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

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

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

International Nuclear Information System (INIS)

Abdou, M.

1984-10-01

The following chapters are included in this study: (1) fusion nuclear issues, (2) survey of experimental needs, (3) requirements of the experiments, (4) non-fusion facilities, (5) fusion facilities for nuclear experiments, and (6) fusion research and development scenarios

VNS: A volumetric neutron source for fusion nuclear technology testing and development

International Nuclear Information System (INIS)

Abdou, M.A.; Peng, Y.K.; Ying, A.Y.

1994-01-01

Recent progress in fusion plasma research and the initiation of the Engineering Design Activity for ITER provide incentives to seriously explore technically sound and logically consistent pathways toward development of fusion as a practical and attractive energy source. A critical goal is the successful construction and operation of a fusion power demonstration plant (DEMO). Major world program strategies call for DEMO operation by the year 2025. Such a date is important in order for fusion to play a significant role in the energy supply market in the second half of the twenty-first century. Without such a DEMO goal, it will be very hard to justify major financial commitments in the near term for major projects such as ITER. The major question is whether a DEMO goal by the year 2025 is attainable from a technical standpoint. This has been the central question being addressed in a study, called VENUS. Results to date show that a DEMO by the year 2025 can be realized if three major facilities begin operation in parallel by the year 2005. These facilities are: (1) ITER, (2) VNS, and (3) IFMIF. Results show that VNS is a necessary element toward DEMO in a strategy consistent with present world program plans. The key requirements to test and develop fusion nuclear components (e.g. blanket) are 1 MW/m 2 neutron wall load, >10 m 2 of test area at the first wall, steady state or long burn plasma operation, fluence of ∼6MWy/m 2 at the first wall in ∼10-12 year period, and duty cycle x availability factor of ∼0.3. Results of the study show that an attractive design envelope for VNS that satisfies the nuclear testing and development requirements exists. Within this design envelope, the most attractive design points for VNS appear to be driven plasma (Q∼1) in tokamak configuration with normal toroidal-field copper coils, major radius 1.5-2.0m, fusion power ∼100MW, and neutron wall load ∼1.5MW/m 2

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.

Accelerators for Fusion Materials Testing

Science.gov (United States)

Knaster, Juan; Okumura, Yoshikazu

Fusion materials research is a worldwide endeavor as old as the parallel one working toward the long term stable confinement of ignited plasma. In a fusion reactor, the preservation of the required minimum thermomechanical properties of the in-vessel components exposed to the severe irradiation and heat flux conditions is an indispensable factor for safe operation; it is also an essential goal for the economic viability of fusion. Energy from fusion power will be extracted from the 14 MeV neutron freed as a product of the deuterium-tritium fusion reactions; thus, this kinetic energy must be absorbed and efficiently evacuated and electricity eventually generated by the conventional methods of a thermal power plant. Worldwide technological efforts to understand the degradation of materials exposed to 14 MeV neutron fluxes >1018 m-2s-1, as expected in future fusion power plants, have been intense over the last four decades. Existing neutron sources can reach suitable dpa (“displacement-per-atom”, the figure of merit to assess materials degradation from being exposed to neutron irradiation), but the differences in the neutron spectrum of fission reactors and spallation sources do not allow one to unravel the physics and to anticipate the degradation of materials exposed to fusion neutrons. Fusion irradiation conditions can be achieved through Li (d, xn) nuclear reactions with suitable deuteron beam current and energy, and an adequate flowing lithium screen. This idea triggered in the late 1970s at Los Alamos National Laboratory (LANL) a campaign working toward the feasibility of continuous wave (CW) high current linacs framed by the Fusion Materials Irradiation Test (FMIT) project. These efforts continued with the Low Energy Demonstrating Accelerator (LEDA) (a validating prototype of the canceled Accelerator Production of Tritium (APT) project), which was proposed in 2002 to the fusion community as a 6.7MeV, 100mA CW beam injector for a Li (d, xn) source to bridge

EPICS - MDSplus integration in the ITER Neutral Beam Test Facility

International Nuclear Information System (INIS)

Luchetta, Adriano; Manduchi, Gabriele; Barbalace, Antonio; Soppelsa, Anton; Taliercio, Cesare

2011-01-01

SPIDER, the ITER-size ion-source test bed in the ITER Neutral Beam Test Facility, is a fusion device requiring a complex central system to provide control and data acquisition, referred to as CODAS. The CODAS software architecture will rely on EPICS and MDSplus, two open-source, collaborative software frameworks, targeted at control and data acquisition, respectively. EPICS has been selected as ITER CODAC middleware and, as the final deliverable of the Neutral Beam Test Facility is the procurement of the ITER Heating Neutral Beam Injector, we decided to adopt this ITER technology. MDSplus is a software package for data management, supporting advanced concepts, such as platform and underlying hardware independence, self description data, and data driven model. The combined use of EPICS and MDSplus is not new in fusion, but their level of integration will be new in SPIDER, achieved by a more refined data access layer. The paper presents the integration software to use effectively EPICS and MDSplus, including the definition of appropriate EPICS records to interact with MDSplus. The MDSplus and EPICS archive concepts are also compared on the basis of performance tests and data streaming is investigated by ad-hoc measurements.

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.

Recent US target-physics-related research in heavy-ion inertial fusion: simulations for tamped targets and for disk experiments in accelerator test facilities

International Nuclear Information System (INIS)

Mark, J.W.K.

1982-01-01

Within the last few years, there have also appeared in the Heavy-Ion Fusion literature several studies of targets which have outer tampers. One-dimensional simulations indicate higher target gains with a judicious amount of tamping. But for these targets, a full investigation has not been carried through in regards to conservative criteria for fluid instabilities as well as reasonable imperfections in target fabrication and illumination symmetry which all affect target ignition and burn. Comparisons of these results with the gain survey of Part I would have to be performed with care. These calculations suggest that experiments relating to high temperature disk heating, as well as beam deposition, focusing and transport can be performed within the context of current design proposals for accelerator test-facilities. Since the test-facilities have lower ion kinetic energy and beam pulse power as compared to reactor drivers, we achieve high-beam intensities at the focal spot by using short focal distance and properly designed beam optics

Scope and status of the USA Engineering Test Facility including relevant TFTR research and development

International Nuclear Information System (INIS)

Becraft, W.R.; Reardon, P.J.

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 progress toward the design and construction of the ETF will reflect the significant achievements of past, present, and future experimental tokamak devices. Some of the features of this foundation of experimental results and relevant engineering designs and operation will derive from the Tokamak Fusion Test Reactor (TFTR) Project, now nearing the completion of its construction phase. 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 (OFE) established the ETF Design Center activity to prepare the design of the ETF. This paper describes the design status of the ETF and discusses some highlights of the TFTR R and D work

Scope and status of the USA Engineering Test Facility including relevant TFTR research and development

International Nuclear Information System (INIS)

Becraft, W.R.; Reardon, P.J.

1981-01-01

The vehicle by which the fusion programme would move into the engineering testing phase of fusion power development is designated the Engineering Test Facility (ETF). The progress toward the design and construction of the ETF will reflect the significant achievements of past, present, and future experimental tokamak devices. Some of the features of this foundation of experimental results and relevant engineering designs and operation will derive from the Tokamak Fusion Test Reactor (TFTR) Project, now nearing the completion of its construction phase. The ETF would provide a test-bed for reactor components in the fusion environment. To initiate preliminary planning for the ETF decision, the Office of Fusion Energy (OFE) established the ETF Design Center activity to prepare the design of the ETF. This paper describes the design status of the ETF and discusses some highlights of the TFTR R and D work. (author)

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

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)

Static and dynamic analyses on the MFTF [Mirror Fusion Test Facility]-B Axicell Vacuum Vessel System: Final report

International Nuclear Information System (INIS)

Ng, D.S.

1986-09-01

The Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory (LLNL) is a large-scale, tandem-mirror-fusion experiment. MFTF-B comprises many highly interconnected systems, including a magnet array and a vacuum vessel. The vessel, which houses the magnet array, is supported by reinforced concrete piers and steel frames resting on an array of foundations and surrounded by a 7-ft-thick concrete shielding vault. The Pittsburgh-Des Moines (PDM) Corporation, which was awarded the contract to design and construct the vessel, carried out fixed-base static and dynamic analyses of a finite-element model of the axicell vessel and magnet systems, including the simulation of various loading conditions and three postulated earthquake excitations. Meanwhile, LLNL monitored PDM's analyses with modeling studies of its own, and independently evaluated the structural responses of the vessel in order to define design criteria for the interface members and other project equipment. The assumptions underlying the finite-element model and the behavior of the axicell vessel are described in detail in this report, with particular emphasis placed on comparing the LLNL and PDM studies and on analyzing the fixed-base behavior with the soil-structure interaction, which occurs between the vessel and the massive concrete vault wall during a postulated seismic event. The structural members that proved sensitive to the soil effect are also reevaluated

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.

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)

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.

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.

Remote handling features of the Fusion Materials Irradiation Test (FMIT) facility

International Nuclear Information System (INIS)

Klos, D.B.; Wierman, R.W.; Kelly, V.P.; Yount, J.A.

1980-01-01

Initial design of the experimental system provided two modes of access to the test cells. The horizontal mode was the predominant one. However, as the design progressed unacceptable risks were identified that increased personnel exposure to radiation and decreased testing availability of the facility. Consequently, vertical-only access was adopted. Remote handling features of both design concepts are described including the technical basis for the transition from the first to the second concept

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

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

Physics and engineering assessments of spherical torus component test facility

International Nuclear Information System (INIS)

Peng, Y.-K.M.; Neumeyer, C.A.; Kessel, C.; Rutherford, P.; Mikkelsen, D.; Bell, R.; Menard, J.; Gates, D.; Schmidt, J.; Synakowski, E.; Grisham, L.; Fogarty, P.J.; Strickler, D.J.; Burgess, T.W.; Tsai, J.; Nelson, B.E.; Sabbagh, S.; Mitarai, O.; Cheng, E.T.; El-Guebaly, L.

2005-01-01

A broadly based study of the fusion engineering and plasma science conditions of a Component Test Facility (CTF), using the Spherical Torus or Spherical Tokamak (ST) configuration, have been carried out. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes Γ n > 4.4x10 13 n/s/cm 2 , over size scales > 10 5 cm 2 and depth scales > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year. The desired chamber conditions can be provided by a CTF with R 0 1.2 m, A = 1.5, elongation ∼ 3.2, I p ∼ 9 MA, B T ∼ 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. Relatively robust ST plasma conditions are adequate, which have been shown achievable [4] without active feedback manipulation of the MHD modes. The ST CTF will test the single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of lowercost CTF devices to suit a variety of fusion engineering science test missions. (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

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

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

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

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

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.

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

Proceedings of the IEA-technical workshop on the test cell system for an international fusion materials irradiation facility, Karlsruhe, Germany, July 3-6, 1995. IEA-implementing agreement for a programme of research and development on fusion materials

International Nuclear Information System (INIS)

Moeslang, A.; Lindau, R.

1995-09-01

After a Conceptual Design Activity (CDA) study on an International Fusion Material Irradiation Facility (IFMIF) has been launched under the auspices of the IEA, working groups and relevant tasks have been defined and agreed in an IEA-workshop that was held September 26-29 1994 at Karlsruhe. For the Test Cell System 11 tasks were identified which can be grouped into the three major fields neutronics, test matrix/users and test cell engineering. In order to discuss recently achieved results and to coordinate necessary activities for an effective design integration, a technical workshop on the Test Cell System was initiated. This workshop was organized on July 3-6 1995 by the Institute for Materials Research I at the Forschungszentrum Karlsruhe and attended by 20 specialists working in the fields neutronics, fusion materials R and D and test cell engineering in the European Union, Japan, and the United States of America. The presentations and discussions during this workshop have shown together with the elaborated lists of action items, that has been achieved in all three fields, and that from the future IFMIF experimental program for a number of materials a database covering widerspread loading conditions up to DEMO-reactor relevant end-of-life damage levels can be expected. (orig.)

Software architecture for the ORNL large coil test facility data system

International Nuclear Information System (INIS)

Blair, E.T.; Baylor, L.R.

1986-01-01

The VAX based data acquisition system for the international fusion superconducting magnetic test facility (IFSMTF) at Oak Ridge National Laboratory (ORNL) is a second generation system that evolved from a PDP-11/60 based system used during the initial phase of facility testing. The VAX based software represents a layered implementation that provides integrated access to all of the data sources within the system, decoupling en-user data retrieval from various front-end data sources through a combination of software architecture and instrumentation data bases. Independent VAX processes manage the various front-end data sources, each being responsible for controlling, monitoring, acquiring, and disposing data and control parameters for access from the data retrieval software

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

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

Gyrotron development at KIT: FULGOR test facility and gyrotron concepts for DEMO

Energy Technology Data Exchange (ETDEWEB)

Schmid, M., E-mail: martin.schmid@kit.edu [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Franck, J.; Kalaria, P.; Avramidis, K.A.; Gantenbein, G.; Illy, S. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Jelonnek, J. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Institute of High Frequency Techniques and Electronics (IHE), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Pagonakis, I. Gr.; Rzesnicki, T. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Thumm, M. [Institute for Pulsed Power and Microwave Technology (IHM), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany); Institute of High Frequency Techniques and Electronics (IHE), Karlsruhe Institute of Technology, Association EURATOM-KIT, Karlsruhe (Germany)

2015-10-15

Highlights: • Substantial extension of the KIT gyrotron test facility FULGOR has started. • FULGOR will be able to test gyrotrons with continuous RF output power up to 4 MW. • Design of 240 GHz gyrotrons for efficient electron cyclotron current drive is progressing. • Output power of 240 GHz gyrotrons with conventional cavity up to 830 kW, with coaxial cavity up to 2 MW is feasible. • Multi-frequency operation with gyrotrons is also possible (170–267 GHz). - Abstract: At the Karlsruhe Institute of Technology (KIT), theoretical and experimental foundations for the development of future gyrotrons for fusion applications are being laid down. This includes the construction of the new Fusion Long Pulse Gyrotron Laboratory (FULGOR) test facility as well as physical design studies towards DEMO-compatible gyrotrons. Initially FULGOR will comprise of a 10 MW CW power supply, a 5 MW water cooling system (upgradeable to 10 MW), a superconducting 10 T magnet, one or two 2 MW ECRH test loads and a new control and data acquisition system for all these elements. The test facility will then be equipped to test the conventional 1 MW or coaxial 2 MW gyrotrons for DEMO, currently under design, as well as possible upgraded gyrotrons for W7-X and ITER. The design of the new high voltage DC power supply (HVDCPS) is flexible enough to handle gyrotrons with 4 MW CW output power (conceivably up to 170 GHz), but also test gyrotrons with higher frequencies (>250 GHz) which, due to physical limitations in the gyrotron design, will require less power but have more stringent demands on voltage stability.

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

The operator interface for the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Lang, N.C.

1986-12-01

The uncertain and most likely changing nature of a large experimental facility like MFTF, as well as its large number of control and monitor points, ruled against the traditional hardware approach involving walls of knobs, dials, oscilloscopes, and strip chart recorders. Rather, from the beginning, project management specified computer control of all systems, and operation of the complete MFTF under an integrated computer control system became a major engineering goal. The Integrated Controls and Diagnostics (ICADS) group was charged with the design and implementation of this control system. We designed a control system with an extremely flexible operator interface which uses computer generated CRT displays for output and pointing devices such as touch sensitive CRT overlays, mice, and joysticks for input. Construction of MFTF was completed at the end of 1985 within the project budget of $241.6M and was followed immediately by a 5 month long acceptance test. During this period (known as PACE test) operators, engineers, and physicists successfully used our computer control system daily to test MFTF. Much of their willingness to forsake the traditional hands-on hardware approach to testing was a result of the powerful and flexible operator interface to the MFTF control system. In this paper, we describe the operator interface with emphasis on the displays, the touch screens, and the mouse. We also report the experiences of users and, in particular, stress those aspects of the user interface they strongly liked and disliked

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

International Nuclear Information System (INIS)

Hosea, J.; Adler, J.H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Ashcroft, D.

1994-09-01

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to ∼9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning; possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS ∼6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored

A stand alone computer system to aid the development of mirror fusion test facility RF heating systems

International Nuclear Information System (INIS)

Thomas, R.A.

1983-01-01

The Mirror Fusion Test Facility (MFTF-B) control system architecture requires the Supervisory Control and Diagnostic System (SCDS) to communicate with a LSI-11 Local Control Computer (LCC) that in turn communicates via a fiber optic link to CAMAC based control hardware located near the machine. In many cases, the control hardware is very complex and requires a sizable development effort prior to being integrated into the overall MFTF-B system. One such effort was the development of the Electron Cyclotron Resonance Heating (ECRH) system. It became clear that a stand alone computer system was needed to simulate the functions of SCDS. This paper describes the hardware and software necessary to implement the SCDS Simulation Computer (SSC). It consists of a Digital Equipment Corporation (DEC) LSI-11 computer and a Winchester/Floppy disk operating under the DEC RT-11 operating system. All application software for MFTF-B is programmed in PASCAL, which allowed us to adapt procedures originally written for SCDS to the SSC. This nearly identical software interface means that software written during the equipment development will be useful to the SCDS programmers in the integration phase

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.

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)

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

Conceptual design of a test facility for the remote handling operations of the ITER Test Blanker Modules

International Nuclear Information System (INIS)

Marqueta, A.; Garcia, I.; Gomez, A.; Garcia, L.; Sedano, E.; Fernandez, I.

2012-01-01

Conceptual Design of a test facility for the remote handling operations of the ITER Test Blanket Modules. Conditions inside a fusion reactor are incompatible with conventional manual maintenance tasks. the same applies for ancillary equipment. As a consequence, it will become necessary to turn to remote visualization and remote handling techniques, which will have in consideration the extreme conditions, both physical and operating, of ITER. Main goal of the project has been the realization of the conceptual design for the test facility for the Test Blanket Modules of ITER and their associated systems, related to the Remote Handling operations regarding the Port Cell area. Besides the definition of the operations and the specification of the main components and ancillary systems of the TBM graphical simulation have been used for the design, verification and validation of the remote handling operations. (Author)

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

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

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

Hybrid testing of lumbar CHARITE discs versus fusions.

Science.gov (United States)

Panjabi, Manohar; Malcolmson, George; Teng, Edward; Tominaga, Yasuhiro; Henderson, Gweneth; Serhan, Hassan

2007-04-20

An in vitro human cadaveric biomechanical study. To quantify effects on operated and other levels, including adjacent levels, due to CHARITE disc implantations versus simulated fusions, using follower load and the new hybrid test method in flexion-extension and bilateral torsion. Spinal fusion has been associated with long-term accelerated degeneration at adjacent levels. As opposed to the fusion, artificial discs are designed to preserve motion and diminish the adjacent-level effects. Five fresh human cadaveric lumbar specimens (T12-S1) underwent multidirectional testing in flexion-extension and bilateral torsion with 400 N follower load. Intact specimen total ranges of motion were determined with +/-10 Nm unconstrained pure moments. The intact range of motion was used as input for the hybrid tests of 5 constructs: 1) CHARITE disc at L5-S1; 2) fusion at L5-S1; 3) CHARITE discs at L4-L5 and L5-S1; 4) CHARITE disc at L4-L5 and fusion at L5-S1; and 5) 2-level fusion at L4-L5-S1. Using repeated-measures single factor analysis of variance and Bonferroni statistical tests (P < 0.05), intervertebral motion redistribution of each construct was compared with the intact. In flexion-extension, 1-level CHARITE disc preserved motion at the operated and other levels, while 2-level CHARITE showed some amount of other-level effects. In contrast, 1- and 2-level fusions increased other-level motions (average, 21.0% and 61.9%, respectively). In torsion, both 1- and 2-level discs preserved motions at all levels. The 2-level simulated fusion increased motions at proximal levels (22.9%), while the 1-level fusion produced no significant changes. In general, CHARITE discs preserved operated- and other-level motions. Fusion simulations affected motion redistribution at other levels, including adjacent levels.

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

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

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

International Fusion Materials Irradiation Facility injector acceptance tests at CEA/Saclay: 140 mA/100 keV deuteron beam characterization

International Nuclear Information System (INIS)

Gobin, R.; Bogard, D.; Chauvin, N.; Chel, S.; Delferrière, O.; Harrault, F.; Mattei, P.; Senée, F.; Cara, P.; Mosnier, A.; Shidara, H.; Okumura, Y.

2014-01-01

In the framework of the ITER broader approach, the International Fusion Materials Irradiation Facility (IFMIF) deuteron accelerator (2 × 125 mA at 40 MeV) is an irradiation tool dedicated to high neutron flux production for future nuclear plant material studies. During the validation phase, the Linear IFMIF Prototype Accelerator (LIPAc) machine will be tested on the Rokkasho site in Japan. This demonstrator aims to produce 125 mA/9 MeV deuteron beam. Involved in the LIPAc project for several years, specialists from CEA/Saclay designed the injector based on a SILHI type ECR source operating at 2.45 GHz and a 2 solenoid low energy beam line to produce such high intensity beam. The whole injector, equipped with its dedicated diagnostics, has been then installed and tested on the Saclay site. Before shipment from Europe to Japan, acceptance tests have been performed in November 2012 with 100 keV deuteron beam and intensity as high as 140 mA in continuous and pulsed mode. In this paper, the emittance measurements done for different duty cycles and different beam intensities will be presented as well as beam species fraction analysis. Then the reinstallation in Japan and commissioning plan on site will be reported

Development of the rf linear accelerator test bed for heavy-ion fusion

International Nuclear Information System (INIS)

Watson, J.M.

1981-01-01

The amount of absorbed energy required by high gain deuterium-tritium targets for inertial confinement fusion reactors is now projected to be greater than 1 Megajoule. It has become apparent that a heavy ion fusion driver is the preferred choice in this scenario. To demonstrate this accelerator-based option, the national program has established two test beds: one at Argonne for the rf linac/storage ring approach, and one at Lawrence Berkeley Laboratory developing an induction linac. The Argonne Beam Development Facility (BDF) would consist of a 40 mA rf linac for Xe + 8 , a storage ring, and a 10 GeV synchrotron. The design and status of the BDF is described as well as future program options to demonstrate as many solutions as possible of the issues involved in this approach

Deuteron beam interaction with Li jet for a neutron source test facility

International Nuclear Information System (INIS)

Hassanein, A.

1995-09-01

Testing and evaluating candidate fusion reactor materials in a high-flux, high-energy neutron environment are critical to the success and economic feasibility of a fusion device. The current understanding of materials behavior in fission-like environments and existing fusion facilities is insufficient to ensure the necessary performance of future fusion reactor components. An accelerator-based deuterium-lithium system to generate the required high neutron flux for material testing is considered to be the most promising approach in the near future. In this system, a high-energy (30-40 MeV) deuteron beam impinges on a high-speed (10-20 m/s) lithium jet to produce the high-energy (>14 MeV) neutrons required to simulate a fusion environment via the Li (d,n) nuclear stripping reaction. Interaction of the high-energy deuteron beam and the subsequent response of the high-speed lithium jet are evaluated in detail. Deposition of the deuteron beam, jet-thermal hydraulic response, lithium-surface vaporization rate, and dynamic stability of the jet are modeled. It is found that lower beam kinetic energies produce higher surface temperature and consequently higher Li vaporization rates. Larger beam sizes significantly reduce both bulk and surface temperatures. Thermal expansion and dynamic velocities (normal to jet direction) due to beam energy deposition and momentum transfer are much lower than jet flow velocity and decrease substantially at lower beam current densities

Deuteron beam interaction with lithium jet in a neutron source test facility

International Nuclear Information System (INIS)

Hassanein, A.

1996-01-01

Testing and evaluating candidate fusion reactor materials in a high-flux, high-energy neutron environment are critical to the success and economic feasibility of a fusion device. The current understanding of materials behavior in fission-like environments and existing fusion facilities is insufficient to ensure the necessary performance of future fusion reactor components. An accelerator-based deuterium-lithium system to generate the required high neutron flux for material testing is considered to be the most promising approach in the near future. In this system, a high-energy (30-40 MeV) deuteron beam impinges on a high-speed (10-20 m/s) lithium jet to produce the high-energy (≥14 MeV) neutrons required to simulate a fusion environment via the Li (d,n) nuclear stripping reaction. Interaction of the high-energy deuteron beam and the subsequent response of the high-speed lithium jet are evaluated in detail. Deposition of the deuteron beam, jet-thermal hydraulic response, lithium-surface vaporization rate, and dynamic stability of the jet are modeled. It is found that lower beam kinetic energies produce higher surface temperature and consequently higher Li vaporization rates. Larger beam sizes significantly reduce both bulk and surface temperatures. Thermal expansion and dynamic velocities (normal to jet direction) due to beam energy deposition and momentum transfer are much lower than jet flow velocity and decrease substantially at lower beam current densities. (orig.)

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

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

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

Software architecture for the ORNL large-coil test facility data system

International Nuclear Information System (INIS)

Blair, E.T.; Baylor, L.R.

1986-01-01

The VAX-based data-acquisition system for the International Fusion Superconducting Magnet Test Facility (IFSMTF) at Oak Ridge National Laboratory (ORNL) is a second-generation system that evolved from a PDP-11/60-based system used during the initial phase of facility testing. The VAX-based software represents a layered implementation that provides integrated access to all of the data sources within the system, decoupling end-user data retrieval from various front-end data sources through a combination of software architecture and instrumentation data bases. Independent VAX processes manage the various front-end data sources, each being responsible for controlling, monitoring, acquiring, and disposing data and control parameters for access from the data retrieval software. This paper describes the software architecture and the functionality incorporated into the various layers of the data system

Software architecture for the ORNL large coil test facility data system

International Nuclear Information System (INIS)

Blair, E.T.; Baylor, L.R.

1986-01-01

The VAX-based data acquisition system for the International Fusion Superconducting Magnet Test Facility (IFSMTF) at Oak Ridge National Laboratory (ORNL) is a second-generation system that evolved from a PDP-11/60-based system used during the initial phase of facility testing. The VAX-based software represents a layered implementation that provides integrated access to all of the data sources within the system, deoupling end-user data retrieval from various front-end data sources through a combination of software architecture and instrumentation data bases. Independent VAX processes manage the various front-end data sources, each being responsible for controlling, monitoring, acquiring and disposing data and control parameters for access from the data retrieval software. This paper describes the software architecture and the functionality incorporated into the various layers of the data system

A Laser Technology Test Facility for Laser Inertial Fusion Energy (LIFE)

International Nuclear Information System (INIS)

Bayramian, A.J.; Campbell, R.W.; Ebbers, C.A.; Freitas, B.L.; Latkowski, J.; Molander, W.A.; Sutton, S.B.; Telford, S.; Caird, J.A.

2010-01-01

A LIFE laser driver needs to be designed and operated which meets the rigorous requirements of the NIF laser system while operating at high average power, and operate for a lifetime of >30 years. Ignition on NIF will serve to demonstrate laser driver functionality, operation of the Mercury laser system at LLNL demonstrates the ability of a diode-pumped solid-state laser to run at high average power, but the operational lifetime >30 yrs remains to be proven. A Laser Technology test Facility (LTF) has been designed to specifically address this issue. The LTF is a 100-Hz diode-pumped solid-state laser system intended for accelerated testing of the diodes, gain media, optics, frequency converters and final optics, providing system statistics for billion shot class tests. These statistics will be utilized for material and technology development as well as economic and reliability models for LIFE laser drivers.

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)

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

The accomplishments of lithium target and test facility validation activities in the IFMIF/EVEDA phase

Science.gov (United States)

Arbeiter, Frederik; Baluc, Nadine; Favuzza, Paolo; Gröschel, Friedrich; Heidinger, Roland; Ibarra, Angel; Knaster, Juan; Kanemura, Takuji; Kondo, Hiroo; Massaut, Vincent; Saverio Nitti, Francesco; Miccichè, Gioacchino; O'hira, Shigeru; Rapisarda, David; Sugimoto, Masayoshi; Wakai, Eiichi; Yokomine, Takehiko

2018-01-01

As part of the engineering validation and engineering design activities (EVEDA) phase for the international fusion materials irradiation facility IFMIF, major elements of a lithium target facility and the test facility were designed, prototyped and validated. For the lithium target facility, the EVEDA lithium test loop was built at JAEA and used to test the stability (waves and long term) of the lithium flow in the target, work out the startup procedures, and test lithium purification and analysis. It was confirmed by experiments in the Lifus 6 plant at ENEA that lithium corrosion on ferritic martensitic steels is acceptably low. Furthermore, complex remote handling procedures for the remote maintenance of the target in the test cell environment were successfully practiced. For the test facility, two variants of a high flux test module were prototyped and tested in helium loops, demonstrating their good capabilities of maintaining the material specimens at the desired temperature with a low temperature spread. Irradiation tests were performed for heated specimen capsules and irradiation instrumentation in the BR2 reactor at SCK-CEN. The small specimen test technique, essential for obtaining material test results with limited irradiation volume, was advanced by evaluating specimen shape and test technique influences.

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

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)

Rocketball Test Facility

Data.gov (United States)

Federal Laboratory Consortium — This test facility offers the capability to emulate and measure guided missile radar cross-section without requiring flight tests of tactical missiles. This facility...

Test and User Facilities | NREL

Science.gov (United States)

Test and User Facilities Test and User Facilities Our test and user facilities are available to | L | M | N | O | P | Q | R | S | T | U | V | W | X | Y | Z B Battery Thermal and Life Test Facility Biochemical Conversion Pilot Plant C Controllable Grid Interface Test System D Dynamometer Test Facilities

CLEAR test facility

CERN Multimedia

Ordan, Julien Marius

2017-01-01

A new user facility for accelerator R&D, the CERN Linear Electron Accelerator for Research (CLEAR), started operation in August 2017. CLEAR evolved from the former CLIC Test Facility 3 (CTF3) used by the Compact Linear Collider (CLIC). The new facility is able to host and test a broad range of ideas in the accelerator field.

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

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

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

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.)

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).

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

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)

MFTF-B PACE tests and final cost report

International Nuclear Information System (INIS)

Krause, K.H.; Kozman, T.A.; Smith, J.L.; Horan, R.J.

1986-10-01

The Mirror Fusion Test Facility (MFTF-B) construction project was successfully completed in February 1986, with the conclusion of the Plant and Capital Equipment (PACE) Tests. This series of tests, starting in September 1985 and running through February 1986, demonstrated the overall machine capabilities and special facilities accomplishments for the Mirror Fusion Test Facility Project

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

Utilization of fission reactors for fusion engineering testing

International Nuclear Information System (INIS)

Deis, G.A.; Miller, L.G.

1985-01-01

Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

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

Tools and methods for implementing the control systems on the Mirror Fusion Test Facility

International Nuclear Information System (INIS)

Minor, E.G.; Labiak, W.G.

1981-01-01

Installation of the major hardware subsystems for MFTF is nearing completion. These subsystems include the Fusion Chamber System, the eighty KV Neutral Beam System, the Superconducting Magnet System, and the Personnel Safety System. The Local Controls group has undertaken a uniform aproach to implementing the control systems for all of these hardware subsystems. This approach has two major aspects: (1) to provide a stand-alone computer control system with a remote, portable terminal so that computer control can be provided at the site of the hardware for initial testing, (2) to provide hardware simulators so that the complicated MFTF computer control system can be tested independent of the hardware. The software and hardware tools which were developed to carry out this plan will be described. Our experiences with bringing up subsystems containing up to 900 separate channels of control and status will also be described

Facilities for technology testing of ITER divertor concepts, models, and prototypes in a plasma environment

International Nuclear Information System (INIS)

Cohen, S.A.

1991-12-01

The exhaust of power and fusion-reaction products from ITER plasma are critical physics and technology issues from performance, safety, and reliability perspectives. Because of inadequate pulse length, fluence, flux, scrape-off layer plasma temperature and density, and other parameters, the present generation of tokamaks, linear plasma devices, or energetic beam facilities are unable to perform adequate technology testing of divertor components, though they are essential contributors to many physics issues such as edge-plasma transport and disruption effects and control. This Technical Requirements Documents presents a description of the capabilities and parameters divertor test facilities should have to perform accelerated life testing on predominantly technological divertor issues such as basic divertor concepts, heat load limits, thermal fatigue, tritium inventory and erosion/redeposition. The cost effectiveness of such divertor technology testing is also discussed

Development of tritium technology for the United States magnetic fusion energy program

International Nuclear Information System (INIS)

Anderson, J.L.; Wilkes, W.R.

1980-01-01

Tritium technology development for the DOE fusion program is taking place principally at three laboratories, Mound Facility, Argonne National Laboratory and the Los Alamos Scientific Laboratory. This paper will review the major aspects of each of the three programs and look at aspects of the tritium technology being developed at other laboratories within the United States. Facilities and experiments to be discussed include the Tritium Effluent Control Laboratory and the Tritium Storage and Delivery System for the Tokamak Fusion Test Reactor at Mound Facility; the Lithium Processing Test Loop and the solid breeder blanket studies at Argonne; and the Tritium Systems Test Assembly at Los Alamos

Potential mirror concepts for radiation testing of fusion reactor materials

International Nuclear Information System (INIS)

Miley, G.H.

1977-01-01

Studies under the University of Illinois PROMETHEUS (Plasma Reactor Optimized for Materials Experimentation for Thermonuclear Energy Usage) project are described that started in 1971 with the realization that a practical fusion-plasma neutron source was feasible with a net-power input (rather than production). The basic objectives were similar to those in later FERF (Fusion Engineering Research Facility) studies: namely, to maximize the neutron flux and usable experimental volume; to include the flexibility to handle a variety of both materials and engineering experiments; to minimize capital and operating costs; and to utilize near- term technology. The PROMETHEUS design provides a neutron flux of approximately 5x10 14 n/cm 2 s by injection of approximately 30 MW of neutral-beams into a 20 cm radius mirror-confined plasma. Charge-exchange bombardment of the first wall is viewed as a key problem in the design and is discussed in some detail. To gain yet higher neutron fluxes for accelerated testing, two alternate designs have been studied: a 'Twin-beam' injection device and a field reversed mirror concept. The latter potentially offers fluxes approaching 10 16 n/cm 2 s but involves more speculative technology. (Auth.)

Status report on the 12T split coil test facility SULTAN

International Nuclear Information System (INIS)

Blau, B.; Aebli, E.; Jakob, B.; Pasztor, G.; Vecsey, G.; della Corte, A.; Pasotti, G.; Sacchetti, N.; Spadoni, M.

1992-01-01

The third phase of upgrading of the superconductor test facility SULTAN into a split coil system (SULTAN III) is in progress. SULTAN III a join project of ENEA (Italy) and PSI (Switzerland) consists of two coil packages, each containing three concentrically mounted superconducting solenoids. Together they will produce a field of nearly 12T between the two coil packages, inside a solenoid bore of 58 cm. The outermost 6T coils have NbTi conductors, whereas the inner 9T and 12T coils are made of A-15 cables. All Nb 3 Sn coils are manufactured by the react-and-wind technique. The split coil arrangement, in connection with a sophisticated sample insert containing a 50 kA superconducting transformer, will allow testing of short samples of high current carrying superconductors, e.g. for fusion applications. The sample insert was designed to allow changing the samples within a few hours without warming up the whole magnet system. This paper deals with the present status and potential of the Split Coil Test Facility SULTAN III

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

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

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)

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.

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.

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)

Test facility for the development of 150-keV, multi-megawatt neutral beam systems

International Nuclear Information System (INIS)

Haughian, W.; Baker, W.R.; Biagi, L.A.; Hopkins, D.B.

1975-11-01

The next generation of CTR experiments, such as the Tokamak Fusion Test Reactor (TFTR), will require neutral-beam injection systems that produce multi-megawatt, 120-keV deuterium-beam pulses of 0.5-second duration. Since present injection systems are operating in the 10- to 40-keV range, an intensive development effort is in progress to meet a 150-keV requirement. The vacuum system and power supplies that make up a test facility to be used in the development of these injectors are described

Integral test of JENDL fusion file

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Integral test of JENDL Fusion File (J-FF) is performed through analyses of available benchmark experiments. As a result, good agreement between the calculated results with J-FF and the measured data is observed as a whole. Thus, J-FF is qualified to be used for nuclear design of fusion reactors. Owing to the high quality evaluation of J-FF, cross section data in J-FF for many nuclides are recommended to be assigned as data in FENDL/E-2.0 in the IAEA Consultants` Meeting held at Karlsruhe, Germany, 24-28 June, 1996. (author)

Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

International Nuclear Information System (INIS)

Murao, Y.; Iguchi, T.; Sugimoto, J.

1988-09-01

The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

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

National Solar Thermal Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The National Solar Thermal Test Facility (NSTTF) is the only test facility in the United States of its type. This unique facility provides experimental engineering...

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

Liquefied Gaseous Fuels Spill Test Facility

International Nuclear Information System (INIS)

1993-02-01

The US Department of Energy's liquefied Gaseous Fuels Spill Test Facility is a research and demonstration facility available on a user-fee basis to private and public sector test and training sponsors concerned with safety aspects of hazardous chemicals. Though initially designed to accommodate large liquefied natural gas releases, the Spill Test Facility (STF) can also accommodate hazardous materials training and safety-related testing of most chemicals in commercial use. The STF is located at DOE's Nevada Test Site near Mercury, Nevada, USA. Utilization of the Spill Test Facility provides a unique opportunity for industry and other users to conduct hazardous materials testing and training. The Spill Test Facility is the only facility of its kind for either large- or small-scale testing of hazardous and toxic fluids including wind tunnel testing under controlled conditions. It is ideally suited for test sponsors to develop verified data on prevention, mitigation, clean-up, and environmental effects of toxic and hazardous gaseous liquids. The facility site also supports structured training for hazardous spills, mitigation, and clean-up. Since 1986, the Spill Test Facility has been utilized for releases to evaluate the patterns of dispersion, mitigation techniques, and combustion characteristics of select materials. Use of the facility can also aid users in developing emergency planning under US P.L 99-499, the Superfund Amendments and Reauthorization Act of 1986 (SARA) and other regulations. The Spill Test Facility Program is managed by the US Department of Energy (DOE), Office of Fossil Energy (FE) with the support and assistance of other divisions of US DOE and the US Government. DOE/FE serves as facilitator and business manager for the Spill Test Facility and site. This brief document is designed to acquaint a potential user of the Spill Test Facility with an outline of the procedures and policies associated with the use of the facility

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.

Structural Test Facility

Data.gov (United States)

Federal Laboratory Consortium — Provides a wide variety of testing equipment, fixtures and facilities to perform both unique aviation component testing as well as common types of materials testing...

Survey of solar thermal test facilities

Energy Technology Data Exchange (ETDEWEB)

Masterson, K.

1979-08-01

The facilities that are presently available for testing solar thermal energy collection and conversion systems are briefly described. Facilities that are known to meet ASHRAE standard 93-77 for testing flat-plate collectors are listed. The DOE programs and test needs for distributed concentrating collectors are identified. Existing and planned facilities that meet these needs are described and continued support for most of them is recommended. The needs and facilities that are suitable for testing components of central receiver systems, several of which are located overseas, are identified. The central contact point for obtaining additional details and test procedures for these facilities is the Solar Thermal Test Facilities Users' Association in Albuquerque, N.M. The appendices contain data sheets and tables which give additional details on the technical capabilities of each facility. Also included is the 1975 Aerospace Corporation report on test facilities that is frequently referenced in the present work.

The role of radiation damage analysis in the fusion program

International Nuclear Information System (INIS)

Doran, D.G.

1983-01-01

The objective of radiation damage analysis is the prediction of the performance of facility components exposed to a radiation environment. The US Magnetic Fusion Energy materials program includes an explicit damage analysis activity within the Damage Analysis and Fundamental Studies (DAFS) Program. Many of the papers in these Proceedings report work done directly or indirectly in support of the DAFS program. The emphasis of this program is on developing procedures, based on an understanding of damage mechanisms, for applying data obtained in diverse radiation environments to the prediction of component behavior in fusion devices. It is assumed that the Fusion Materials Irradiation Test Facility will be available in the late 1980s to test (and calibrate where necessary) correlation procedures to the high fluences expected in commercial reactors. (orig.)

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.

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.)

The Integral Test Facility Karlstein

Directory of Open Access Journals (Sweden)

Stephan Leyer

2012-01-01

Full Text Available The Integral Test Facility Karlstein (INKA test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.

GPS Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Global Positioning System (GPS) Test Facility Instrumentation Suite (GPSIS) provides great flexibility in testing receivers by providing operational control of...

New electron beam facility for irradiated plasma facing materials testing in hot cell

International Nuclear Information System (INIS)

Sakamoto, N.; Kawamura, H.; Akiba, M.

1995-01-01

Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop of plasma facing components which can resist these. Then, we have established electron beam heat facility (open-quotes OHBISclose quotes, Oarai Hot-cell electron Beam Irradiating System) at a hot cell in JMTR (Japan Materials Testing Reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30kV (constant) and 1.7A, respectively. The loading time of electron beam is more than 0.1ms. The shape of vacuum vessel is cylindrical, and the mainly dimensions are 500mm in inner diameter, 1000mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for thermal shock test has been established in a hot cell. And performance estimation on the electron beam is being conducted. Presently, the devices for heat loading tests under steady state will be added to this facility

Overview of principles and challenges of fusion nuclear technology

International Nuclear Information System (INIS)

Abdou, M.

2007-01-01

Fusion offers very attractive features as a sustainable, broadly available energy source: no emissions of greenhouse gases, no risk of severe accident, and no long-lived radioactive waste. Significant advances in the science and technology of fusion have been realized in the past decades. Seven countries (EU, Japan, USA, Russia, S. Korea, China, and India) comprising about half the world population are constructing a major magnetic fusion facility, called ITER, in France. The objectives of ITER are to demonstrate self-sustaining burning fusion plasma and to test fusion technologies relevant to fusion reactor. Many challenges to the practical utilization of fusion energy remain ahead. Among these challenges is the successful development of Fusion Nuclear Technology (FNT). FNT includes those fusion system components circumscribing the plasma and responsible for tritium production and processing, heat removal at high temperature and power density, and high heat flux components. FNT components face a new and more challenging environment than experienced by any previous nuclear application. Beyond plasma physics, FNT has most of the remaining feasibility and attractiveness issues in the development of fusion as an energy source. The blanket, a key FNT component, determines the critical path to DEMO. The blanket is exposed to an intense radiation environment. Radioactivity and decay heat can be produced in the structure and other blanket elements. Hence, material choices have a large impact on safety and environmental attractiveness. The unique conditions of the fusion environment include high radiation flux, high surface heat flux, strong 3-D-component magnetic field with large gradients, and ultra-low vacuum. These conditions, together with the requirements for high-temperature operation and tritium self-sufficiency, make blanket design and development challenging tasks. The blanket concepts being considered worldwide can be classified into solid breeders and liquid

The ICRH tokamak fusion test reactor

International Nuclear Information System (INIS)

Perkins, F.W.

1976-01-01

A Tokamak Fusion Test Reactor where the ion are maintained at Tsub(i) approximately 20keV>Tsub(e) approximately 7keV by ion-cyclotron resonance heating is shown to produce an energy amplification of Q>2 provided the principal ion energy loss channel is via collisional transfer to the electrons. Such a reactor produces 19MW of fusion power to the electrons. Such a reactor produces 19MW of fusion power and requires a 50MHz radio-frequency generator capable of 50MW peak power; it is otherwise compatible with the conceptual design for the Princeton TFTR. The required n tausub(E) values for electrons and ions are respectively ntausub(Ee)>1.5.10 13 cm -3 -sec and ntausub(Ei)>4.10 13 cm -3 -sec. The principal areas where research is needed to establish this concept are: tokamak transport calculations, ICRH physics, trapped-particle instability energy losses, tokamak equilibria with high values of βsub(theta), and, of course, impurities

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

Energy Systems Test Area (ESTA). Power Systems Test Facilities

Science.gov (United States)

Situ, Cindy H.

2010-01-01

This viewgraph presentation provides a detailed description of the Johnson Space Center's Power Systems Facility located in the Energy Systems Test Area (ESTA). Facilities and the resources used to support power and battery systems testing are also shown. The contents include: 1) Power Testing; 2) Power Test Equipment Capabilities Summary; 3) Source/Load; 4) Battery Facilities; 5) Battery Test Equipment Capabilities Summary; 6) Battery Testing; 7) Performance Test Equipment; 8) Battery Test Environments; 9) Battery Abuse Chambers; 10) Battery Abuse Capabilities; and 11) Battery Test Area Resources.

Corrosion Testing Facility

Data.gov (United States)

Federal Laboratory Consortium — The Corrosion Testing Facility is part of the Army Corrosion Office (ACO). It is a fully functional atmospheric exposure site, called the Corrosion Instrumented Test...

Ballistic Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Ballistic Test Facility is comprised of two outdoor and one indoor test ranges, which are all instrumented for data acquisition and analysis. Full-size aircraft...

Distributed Energy Resources Test Facility

Data.gov (United States)

Federal Laboratory Consortium — NREL's Distributed Energy Resources Test Facility (DERTF) is a working laboratory for interconnection and systems integration testing. This state-of-the-art facility...

Structural materials for fusion and spallation sources

International Nuclear Information System (INIS)

Cottrell, G.A.; Baker, L.J.

2003-01-01

Experimental investigation of neutron-induced irradiation damage in structural materials is fundamental to the development of magnetic confinement fusion. Proposals for the testing of candidate materials are described, indicating that a period of at least 10 years will elapse before a suitable high neutron fluence fusion test facility becomes available. In this circumstance, the possibility that neutron spallation sources could be exploited to shorten the time-scale of fusion materials development is attractive. Although fusion displacement and transmutation reaction rates can be replicated in spallation sources, there are significant differences arising from the harder neutron spectra and the presence of energetic protons. These differences, including higher energy PKA, electron heating effects, transmutation rates and pulsing are described and their consequences discussed, together with the concomitant development of theoretical models, needed to understand the effects. It is concluded that spallation source experiments could make a significant contribution to the database required for the validation of theoretical models, and hence reduce the time scale of fusion materials development

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)

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.)

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

Impurity Control Test Facility (ICTF) for the study of fusion reactor plasma/edge materials interactions

International Nuclear Information System (INIS)

Brooks, J.N.; Mattas, R.F.; Ehst, D.A.; Boley, C.D.; Hershkowitz, N.

1984-05-01

A test facility for investigating many of the impurity control issues associated with the interactions of materials with the plasma edge is outlined. Analysis indicates that the plasma edge conditions expected in TFCX, INTOR, etc. can be readily produced at the end cells of an rf stabilized mirror, similar in some respects to the Phaedrus device at the University of Wisconsin. A steady-state, Impurity Control Test Facility (ICTF) based on such a mirror device is expected to produce a plasma with typical parameters of n/sub e/ approx. 3 x 10 18 m -3 , T/sub e/ = 50 eV, and T/sub i/ = 100 eV at each end cell. A heat load of approx. 2 MW/m 2 over areas of approx. 1600 cm 2 could be produced at each end with 800 kW of ICRH power. These conditions would provide a unique capability for examining issues such as erosion/redeposition behavior, properties of redeposited materials, high recycling regimes, plasma edge operating limits for high-Z materials, and particle pumping efficiencies for limiter and divertor designs

Accelerator development for heavy ion fusion

International Nuclear Information System (INIS)

Talbert, W.L. Jr.; Sawyer, G.A.

1980-01-01

Accelerator technology development is presented for heavy ion drivers used in inertial confinement fusion. The program includes construction of low-velocity ''test bed'' accelerator facilities, development of analytical and experimental techniques to characterize ion beam behavior, and the study of ion beam energy deposition

Aircraft Test & Evaluation Facility (Hush House)

Data.gov (United States)

Federal Laboratory Consortium — The Aircraft Test and Evaluation Facility (ATEF), or Hush House, is a noise-abated ground test sub-facility. The facility's controlled environment provides 24-hour...

Availability of fusion plants employing a Monte Carlo simulation computer code

International Nuclear Information System (INIS)

Musicki, Z.

1984-01-01

The fusion facilities being built or designed will have availability problems due to their complexity and employment of not yet fully developed technologies. Low availability of test facilities will have an adverse impact on the learning time and will therefore push back the commercialization date of fusion. Low availability of commercial electric power plants will increase the cost of electricity and make fusion a less-attractive power source. Thus, the time to study the availability problems of fusion plants and suggest improvements is now, before costly mistakes are committed. This study is an initial effort in the area and is an attempt to develop methods for calculation of system's performance, specifically its availability, start collecting necessary data and identify the areas where data are lacking, as well as to point out the subsystems where resources need to be applied in order to bring about an acceptable system performance. The method used to study availability is a simulation computer code based on the Monte Carlo process and developed by the author. The fusion systems analyzed were TASKA (a tandem mirror test facility design) and MARS (a tandem mirror power plant design). The model and available data were employed to find that the most critical subsystems needing further work are the neutral beams, RF heating subsystems, direct convertor, and certain magnets

Manual for operation of the multipurpose thermalhydraulic test facility TOPFLOW (Transient Two Phase Flow Test Facility)

International Nuclear Information System (INIS)

Beyer, M.; Carl, H.; Schuetz, H.; Pietruske, H.; Lenk, S.

2004-07-01

The Forschungszentrum Rossendorf (FZR) e. V. is constructing a new large-scale test facility, TOPFLOW, for thermalhydraulic single effect tests. The acronym stands for transient two phase flow test facility. It will mainly be used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of computational fluid dynamic (CFD) codes. The manual of the test facility must always be available for the staff in the control room and is restricted condition during operation of personnel and also reconstruction of the facility. (orig./GL)

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

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

Conceptual design study of a scyllac fusion test reactor

International Nuclear Information System (INIS)

Thomassen, K.I.

1975-07-01

The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements

Eccentric Coil Test Facility (ECTF)

International Nuclear Information System (INIS)

Burn, P.B.; Walstrom, P.L.; Anderson, W.C.; Marguerat, E.F.

1975-01-01

The conceptual design of a facility for testing superconducting coils under some conditions peculiar to tokamak systems is given. A primary element of the proposed facility is a large 25 MJ background solenoid. Discussions of the mechanical structure, the stress distribution and the thermal stability for this coil are included. The systems for controlling the facility and diagnosing test coil behavior are also described

Ouellette Thermal Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Thermal Test Facility is a joint Army/Navy state-of-the-art facility (8,100 ft2) that was designed to:Evaluate and characterize the effect of flame and thermal...

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)

Engine Test Facility (ETF)

Data.gov (United States)

Federal Laboratory Consortium — The Air Force Arnold Engineering Development Center's Engine Test Facility (ETF) test cells are used for development and evaluation testing of propulsion systems for...

IFMIF suitability for evaluation of fusion functional materials

International Nuclear Information System (INIS)

Casal, N.; Sordo, F.; Mota, F.; Jordanova, J.; Garcia, A.; Ibarra, A.; Vila, R.; Rapisarda, D.; Queral, V.; Perlado, M.

2011-01-01

The International Fusion Materials Irradiation Facility (IFMIF) is a future neutron source based on the D-Li stripping reaction, planned to test candidate fusion materials at relevant fusion irradiation conditions. During the design of IFMIF special attention was paid to the structural materials for the blanket and first wall, because they will be exposed to the most severe irradiation conditions in a fusion reactor. Also the irradiation of candidate materials for solid breeder blankets is planned in the IFMIF reference design. This paper focuses on the assessment of the suitability of IFMIF irradiation conditions for testing functional materials to be used in liquid blankets and diagnostics systems, since they are been also considered within IFMIF objectives. The study has been based on the analysis and comparison of the main expected irradiation parameters in IFMIF and DEMO reactor.

New electron beam facility for irradiated plasma facing materials testing in hot cell

International Nuclear Information System (INIS)

Shimakawa, S.; Akiba, M.; Kawamura, H.

1996-01-01

Since plasma facing components such as the first wall and the divertor for the next step fusion reactors are exposed to high heat loads and high energy neutron flux generated by the plasma, it is urgent to develop plasma facing components which can resist these. We have established electron beam heat facility ('OHBIS', Oarai hot-cell electron beam irradiating system) at a hot cell in JMTR (Japan materials testing reactor) hot laboratory in order to estimate thermal shock resistivity of plasma facing materials and heat removal capabilities of divertor elements under steady state heating. In this facility, irradiated plasma facing materials (beryllium, carbon based materials and so on) and divertor elements can be treated. This facility consists of an electron beam unit with the maximum beam power of 50 kW and the vacuum vessel. The acceleration voltage and the maximum beam current are 30 kV (constant) and 1.7 A, respectively. The loading time of the electron beam is more than 0.1 ms. The shape of vacuum vessel is cylindrical, and the main dimensions are 500 mm in inside diameter, 1000 mm in height. The ultimate vacuum of this vessel is 1 x 10 -4 Pa. At present, the facility for the thermal shock test has been established in a hot cell. The performance of the electron beam is being evaluated at this time. In the future, the equipment for conducting static heat loadings will be incorporated into the facility. (orig.)

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.

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.

Several new thermo-hydraulic test facilities in NPIC

International Nuclear Information System (INIS)

Ye Shurong; Sun Yufa; Ji Fuyun; Zong Guifang; Guo Zhongchuan

1997-01-01

Several new thermo-hydraulic test facilities are under construction in Nuclear Power Institute of Chinese (NPIC) at Chengdu. These facilities include: 1. Nuclear Power Component Comprehensive Test Facility. 2. Reactor Hydraulic Modeling Test Facility. 3. Control Rod Drive Line Hydraulic Test Facility. 4. Large Scale Thermo-Hydraulic Test Facility. The construction of these facilities will make huge progress in the research and development capability of nuclear power technology in CHINA. The author will present a brief description of the design parameters flowchart and test program of these facilities

Cryogenic aspects of the mirror fusion test facility

International Nuclear Information System (INIS)

Sterbentz, W.H.; Nelson, R.L.

1979-01-01

This paper covers the design and construction of the MFTF cryogenic system and a description of the operating procedures throughout the many functional modes. The coils and the cryopanels for maintaining the high vacuum environment weigh 417,000 kg (920,000 lb) and must be cooled from room temperature to 4.5 k. The cryogenic system for MFTF consists of a closed-loop helium system with a 3000-W helium refrigerator that uses gas-bearing expansion turbines and oil-flooded screw compressors. In addition, liquid helium storage facilities have adequate capacity for standby operation, and a complete helium-purification plant is capable of processing 17 m 3 /min (600 scfm). An open-loop liquid nitrogen system (with provision for later addition of a nitrogen recondenser) provides the required refrigeration for the radiation shields that must be maintained at 85 K

Integrated Approach to Dense Magnetized Plasmas Applications in Nuclear Fusion Technology. Report of a Coordinated Research Project 2007-2011

International Nuclear Information System (INIS)

2013-04-01

Through its coordinated research activities, the IAEA promotes the development and application of nuclear technologies in Member States. The scientific and technical knowledge required for the construction and operation of large nuclear fusion research facilities, including ITER and the Laser Megajoule in France, and the Z machine and the National Ignition Facility in the United States of America, necessitates several accompanying research and development programmes in physics and technology. This is particularly true in the areas of materials science and fusion technology. Hence, the long standing IAEA effort to conduct coordinated research projects (CRPs) in these areas is aimed at: (i) the development of appropriate technical tools to investigate the issue of materials damage and degradation in a fusion plasma environment; and (ii) the emergence of a knowledge based understanding of the various processes underlying materials damage and degradation, thereby leading to the identification of suitable candidate materials fulfilling the stringent requirements of a fusion environment in any next step facility. Dense magnetized plasma (DMP) devices serve as a first test bench for testing of fusion relevant plasma facing materials, diagnostic development and calibration, technologies and scaling to conceptual principles of larger devices while sophisticated testing facilities such as the International Fusion Materials Irradiation Facility (IFMIF) are being designed. The CRP on Integrated Approach to Dense Magnetized Plasmas Applications in Nuclear Fusion Technology described herein was initiated in 2007 with the participation of 12 research institutions in 8 Member States and was concluded in 2011. It was designed with specific research objectives falling into two main categories: support to mainstream fusion research and development of DMP technology. This publication is a compilation of the individual reports submitted by the 12 CRP participants. These reports discuss

Computer-based data acquisition system in the Large Coil Test Facility

International Nuclear Information System (INIS)

Gould, S.S.; Layman, L.R.; Million, D.L.

1983-01-01

The utilization of computers for data acquisition and control is of paramount importance on large-scale fusion experiments because they feature the ability to acquire data from a large number of sensors at various sample rates and provide for flexible data interpretation, presentation, reduction, and analysis. In the Large Coil Test Facility (LCTF) a Digital Equipment Corporation (DEC) PDP-11/60 host computer with the DEC RSX-11M operating system coordinates the activities of five DEC LSI-11/23 front-end processors (FEPs) via direct memory access (DMA) communication links. This provides host control of scheduled data acquisition and FEP event-triggered data collection tasks. Four of the five FEPs have no operating system

Developments on the RF system for the Fusion Materials Irradiation Test Facility accelerator

International Nuclear Information System (INIS)

Fazio, M.V.; Johnson, H.P.; Riggin, D.M.

1979-01-01

The rf system for the Fusion Materials Irradiation Test (FMIT) accelerator is currently in the design phase at the Los Alamos Scientific Laboratory (LASL). The 35-MeV, 100-mA deuteron beam will require approximately 6 MW of rf power at 80 MHz. The EIMAC 8973 power tetrode, capable of a 600-kW cw output, has been chosen as the final amplifier tube for each of 15 amplifier chains. The final power stage of each chain is designed to perform as a linear Class B amplifier. Each low-power rf system (less than or equal to 100W) is to be phase, amplitude, and frequency controlled to provide a drive signal for each high-power amplifier. Beam dynamics for particle acceleration and for minimal beam spill require each rf amplifier output to be phase controlled to +-1 0 . The amplitude of the accelerating field must be held to +-1%. A varactor-tuned electronic phase shifter and a linear phase detector are under development for use in this system. To complement hardware development, analog computer simulations are being performed to optimize the closed-loop control characteristics of the system

Dosimetry program for characterization of the FMIT facility

International Nuclear Information System (INIS)

Gold, R.; McElroy, W.N.; Fuller, J.L.; Sheen, E.M.; Dierckx, R.

1979-01-01

The environmental characterization program for the Fusion Materials Irradiaton Test (FMIT) facility is presented. Requirements for the development and testing of Magnetic Fusion Energy (MFE) materials together with the complexity of the FMIT (d,Li) generated radiation field warrant a multifaceted dosimetric approach. Specific passive, active and calculational dosimetry efforts comprising this multifaceted approach are described. Special emphasis is given to those dosimetry capabilities uniquely required to characterize FMIT

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)

Cryogenic test facility at VECC, Kolkata

International Nuclear Information System (INIS)

Sarkar, Amit; Bhunia, Uttam; Pradhan, J.; Sur, A.; Bhandari, R.K.; Ranganathan, R.

2003-01-01

In view of proposed K-500 superconducting cyclotron project, cryogenic test facility has been set up at the centre. The facility can broadly be categorized into two- a small scale test facility and a large scale test facility. This facility has been utilized for the calibration of liquid helium level probe, cryogenic temperature probe, and I-B plot for a 7 T superconducting magnet. Spiral-shaped superconducting short sample with specific dimension and specially designed stainless steel sample holder has already been developed for the electrical characterisation. The 1/5 th model superconducting coil along with its quench detection circuit and dump resistor has also been developed

Chemical analysis developments for fusion materials studies

International Nuclear Information System (INIS)

McCown, J.J.; Baldwin, D.L.; Keough, R.F.; Van der Cook, B.P.

1985-04-01

Several projects at Hanford under the management of the Westinghouse Hanford Company have involved research and development (R and D) on fusion materials. They include work on the Fusion Materials Irradiation Test Facility and its associated Experimental Lithium System; testing of irradiated lithium compounds as breeding materials; and testing of Li and Li-Pb alloy reactions with various atmospheres, concrete, and other reactor materials for fusion safety studies. In the course of these projects, a number of interesting and challenging analytical chemistry problems were encountered. They include sampling and analysis of lithium while adding and removing elements of interest; sampling, assaying and compound identification efforts on filters, aerosol particles and fire residues; development of dissolution and analysis techniques for measuring tritium and helium in lithium ceramics including oxides, aluminates, silicates and zirconates. An overview of the analytical chemistry development problems plus equipment and procedures used will be presented

Tandem mirror technology demonstration facility

Energy Technology Data Exchange (ETDEWEB)

1983-10-01

This report describes a facility for generating engineering data on the nuclear technologies needed to build an engineering test reactor (ETR). The facility, based on a tandem mirror operating in the Kelley mode, could be used to produce a high neutron flux (1.4 MW/M/sup 2/) on an 8-m/sup 2/ test area for testing fusion blankets. Runs of more than 100 h, with an average availability of 30%, would produce a fluence of 5 mW/yr/m/sup 2/ and give the necessary experience for successful operation of an ETR.

Tandem mirror technology demonstration facility

International Nuclear Information System (INIS)

1983-10-01

This report describes a facility for generating engineering data on the nuclear technologies needed to build an engineering test reactor (ETR). The facility, based on a tandem mirror operating in the Kelley mode, could be used to produce a high neutron flux (1.4 MW/M 2 ) on an 8-m 2 test area for testing fusion blankets. Runs of more than 100 h, with an average availability of 30%, would produce a fluence of 5 mW/yr/m 2 and give the necessary experience for successful operation of an ETR

Construction and commissioning test report of the CEDM test facility

Energy Technology Data Exchange (ETDEWEB)

Chung, C. H.; Kim, J. T.; Park, W. M.; Youn, Y. J.; Jun, H. G.; Choi, N. H.; Park, J. K.; Song, C. H.; Lee, S. H.; Park, J. K

2001-02-01

The test facility for performance verification of the control element drive mechanism (CEDM) of next generation power plant was installed at the site of KAERI. The CEDM was featured a mechanism consisting of complicated mechanical parts and electromagnetic control system. Thus, a new CEDM design should go through performance verification tests prior to it's application in a reactor. The test facility can simulate the reactor operating conditions such as temperature, pressure and water quality and is equipped with a test chamber to accomodate a CEDM as installed in the power plant. This test facility can be used for the following tests; endurance test, coil cooling test, power measurement and reactivity rod drop test. The commissioning tests for the test facility were performed up to the CEDM test conditions of 320 C and 150 bar, and required water chemistry was obtained by operating the on-line water treatment system.

Construction and commissioning test report of the CEDM test facility

International Nuclear Information System (INIS)

Chung, C. H.; Kim, J. T.; Park, W. M.; Youn, Y. J.; Jun, H. G.; Choi, N. H.; Park, J. K.; Song, C. H.; Lee, S. H.; Park, J. K.

2001-02-01

The test facility for performance verification of the control element drive mechanism (CEDM) of next generation power plant was installed at the site of KAERI. The CEDM was featured a mechanism consisting of complicated mechanical parts and electromagnetic control system. Thus, a new CEDM design should go through performance verification tests prior to it's application in a reactor. The test facility can simulate the reactor operating conditions such as temperature, pressure and water quality and is equipped with a test chamber to accomodate a CEDM as installed in the power plant. This test facility can be used for the following tests; endurance test, coil cooling test, power measurement and reactivity rod drop test. The commissioning tests for the test facility were performed up to the CEDM test conditions of 320 C and 150 bar, and required water chemistry was obtained by operating the on-line water treatment system

Ice condenser testing facility and plans

International Nuclear Information System (INIS)

Kannberg, L.D.; Ross, B.A.; Eschbach, E.J.; Ligotke, M.W.

1987-01-01

A facility is being constructed to experimentally validate the ICEDF computer code. The code was developed to estimate the extent of fission product retention in the ice compartments of pressurized water reactor ice condenser containment systems during severe accidents. The design and construction of the facility is based on a test design that addresses the validation needs of the code for conditions typical of those expected to occur during severe pressurized water reactor accidents. Detailed facility design has followed completion of a test design (i.e., assembled test cases each involving a different set of aerosol and thermohydraulic flow conditions). The test design was developed with the aid of statistical test design software and was scrutinized for applicability with the aid of ICEDF simulations. The test facility will incorporate a small section of a prototypic ice condenser (e.g., a cross section comprising the equivalent of four 1-ft-diameter ice baskets to their full prototypic height of 48 ft). The development of the test design, the detailed facility design, and the construction progress are described in this paper

Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion with Validation from Mechanical Testing

International Nuclear Information System (INIS)

Anderson, Michael T.; Cinson, Anthony D.; Crawford, Susan L.; Doctor, Steven R.; Moran, Traci L.; Watts, Michael W.

2010-01-01

Studies at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, are being conducted to evaluate nondestructive examinations (NDE) coupled with mechanical testing of butt fusion joints in high-density polyethylene (HDPE) pipe for assessing lack of fusion. The work provides information to the U.S. Nuclear Regulatory Commission (NRC) on the effectiveness of volumetric inspection techniques of HDPE butt fusion joints in Section III, Division 1, Class 3, buried piping systems in nuclear power plants. This paper describes results from preliminary assessments using ultrasonic and microwave nondestructive techniques and mechanical testing with the high-speed tensile impact test and the side-bend test for determining joint integrity. A series of butt joints were fabricated in 3408, 12-in. IPS DR-11 HDPE material by varying the fusion parameters to create good joints and joints containing a range of lack-of-fusion conditions. Six of these butt joints were volumetrically examined with time-of-flight diffraction (TOFD), phased-array (PA) ultrasound, and the Evisive microwave system. The outer-diameter weld beads were removed for the microwave inspection. In two of the four pipes, both the outer and inner weld beads were removed and the pipe joints re-evaluated. The pipes were sectioned and the joints destructively evaluated with the side-bend test by cutting portions of the fusion joint into slices that were planed and bent. The last step in this limited study will be to correlate the fusion parameters, nondestructive, and destructive evaluation results to validate the effectiveness of what each NDE technology detects and what each does not detect. The results of the correlation will be used in identifying any future work that is needed.

The PIREX proton irradiation facility

Energy Technology Data Exchange (ETDEWEB)

Victoria, M. [Association EURATOM, Villigen (Switzerland)

1995-10-01

The proton Irradiation Experiment (PIREX) is a materials irradiation facility installed in a beam line of the 590 MeV proton accelerator at the Paul Scherrer Institute. Its main purpose is the testing of candidate materials for fusion reactor components. Protons of this energy produce simultaneously displacement damage and spallation products, amongst them helium and can therefore simulate any possible synergistic effects of damage and helium, that would be produced by the fusion neutrons.

The PIREX proton irradiation facility

International Nuclear Information System (INIS)

Victoria, M.

1995-01-01

The proton Irradiation Experiment (PIREX) is a materials irradiation facility installed in a beam line of the 590 MeV proton accelerator at the Paul Scherrer Institute. Its main purpose is the testing of candidate materials for fusion reactor components. Protons of this energy produce simultaneously displacement damage and spallation products, amongst them helium and can therefore simulate any possible synergistic effects of damage and helium, that would be produced by the fusion neutrons

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)

Mark 1 Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Mark I Test Facility is a state-of-the-art space environment simulation test chamber for full-scale space systems testing. A $1.5M dollar upgrade in fiscal year...

Pavement Testing Facility

Data.gov (United States)

Federal Laboratory Consortium — Comprehensive Environmental and Structural AnalysesThe ERDC Pavement Testing Facility, located on the ERDC Vicksburg campus, was originally constructed to provide an...

Nuclear irradiation parameters of beryllium under fusion, fission and IFMIF irradiation conditions

International Nuclear Information System (INIS)

Fischer, U.; Chen, Y.; Leichtle, D.; Simakov, S.; Moeslang, A.; Vladimirov, P.

2004-01-01

A computational analysis is presented of the nuclear irradiation parameters for Beryllium under irradiation in typical neutron environments of fission and fusion reactors, and of the presently designed intense fusion neutron source IFMIF. The analysis shows that dpa and Tritium production rates at fusion relevant levels can be achieved with existing high flux fission reactors while the achievable Helium production is too low. The resulting He-Tritium and He/dpa ratios do not meet typical fusion irradiation conditions. Irradiation simulations in the medium flux test modules of the IFMIF neutron source facility were shown to be more suitable to match fusion typical irradiation conditions. To achieve sufficiently high production rates it is suggested to remove the creep-fatigue testing machine together with the W spectra shifter plate and move the tritium release module upstream towards the high flux test module. (author)

Hot Hydrogen Test Facility

International Nuclear Information System (INIS)

W. David Swank

2007-01-01

The core in a nuclear thermal rocket will operate at high temperatures and in hydrogen. One of the important parameters in evaluating the performance of a nuclear thermal rocket is specific impulse, ISp. This quantity is proportional to the square root of the propellant's absolute temperature and inversely proportional to square root of its molecular weight. Therefore, high temperature hydrogen is a favored propellant of nuclear thermal rocket designers. Previous work has shown that one of the life-limiting phenomena for thermal rocket nuclear cores is mass loss of fuel to flowing hydrogen at high temperatures. The hot hydrogen test facility located at the Idaho National Lab (INL) is designed to test suitability of different core materials in 2500 C hydrogen flowing at 1500 liters per minute. The facility is intended to test non-uranium containing materials and therefore is particularly suited for testing potential cladding and coating materials. In this first installment the facility is described. Automated Data acquisition, flow and temperature control, vessel compatibility with various core geometries and overall capabilities are discussed

Textiles Performance Testing Facilities

Data.gov (United States)

Federal Laboratory Consortium — The Textiles Performance Testing Facilities has the capabilities to perform all physical wet and dry performance testing, and visual and instrumental color analysis...

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 ICF targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effect 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

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

Superconducting magnets for fusion applications

International Nuclear Information System (INIS)

Henning, C.D.

1987-01-01

Fusion magnet technology has made spectacular advances in the past decade; to wit, the Mirror Fusion Test Facility and the Large Coil Project. However, further advances are still required for advanced economical fusion reactors. Higher fields to 14 T and radiation-hardened superconductors and insulators will be necessary. Coupled with high rates of nuclear heating and pulsed losses, the next-generation magnets will need still higher current density, better stability and quench protection. Cable-in-conduit conductors coupled with polyimide insulations and better steels seem to be the appropriate path. Neutron fluences up to 10 19 neutrons/cm 2 in niobium tin are achievable. In the future, other amorphous superconductors could raise these limits further to extend reactor life or decrease the neutron shielding and corresponding reactor size

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

Millimeter-wave Instrumentation Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Millimeter-wave Instrumentation Test Facility conducts basic research in propagation phenomena, remote sensing, and target signatures. The facility has a breadth...

Ground test facility for nuclear testing of space reactor subsystems

International Nuclear Information System (INIS)

Quapp, W.J.; Watts, K.D.

1985-01-01

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

Environmental Test Facility (ETF)

Data.gov (United States)

Federal Laboratory Consortium — The Environmental Test Facility (ETF) provides non-isolated shock testing for stand-alone equipment and full size cabinets under MIL-S-901D specifications. The ETF...

Safety overview of the National Ignition Facility

International Nuclear Information System (INIS)

Brereton, S.J.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.

1996-01-01

The National Ignition Facility (NIF) is a proposed US Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium- tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF. It provides an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low

40 CFR 792.31 - Testing facility management.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 31 2010-07-01 2010-07-01 true Testing facility management. 792.31... facility management. For each study, testing facility management shall: (a) Designate a study director as... appropriately tested for identity, strength, purity, stability, and uniformity, as applicable. (e) Assure that...

40 CFR 160.31 - Testing facility management.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Testing facility management. 160.31... GOOD LABORATORY PRACTICE STANDARDS Organization and Personnel § 160.31 Testing facility management. For each study, testing facility management shall: (a) Designate a study director as described in § 160.33...

EDITORIAL: Safety aspects of fusion power plants

Science.gov (United States)

Kolbasov, B. N.

2007-07-01

&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 behaviour 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 and improving fusion facility design in terms of accident safety, worker safety, and waste disposal. There are three papers considering safety issues of the test blanket modules (TBM) producing tritium to be installed in ITER. These modules represent different concepts of demonstration fusion power facilities (DEMO). L. Boccaccini et al (Germany) analyses the possibility of jeopardizing the ITER safety under specific accidents in the European helium-cooled pebble-bed TBM, e.g. pressurization of the vacuum vessel (VV), hydrogen production from the Be-steam reaction, the possible interconnection between the port cell and VV causing air ingress. Safety analysis is also presented for Chinese TBM with a helium-cooled solid breeder to be tested in ITER by Z. Chen et al (China). Radiological inventories, afterheat, waste disposal ratings, electromagnetic characteristics, LOCA and tritium safety management are considered. An overview of a preliminary safety analysis performed for a US proposed TBM is presented by B. Merrill et al (USA). This DEMO relevant dual coolant liquid lead-lithium TBM has been explored both in the USA and EU. T. Pinna et al (Italy) summarize the six-year development of a failure rate database for fusion specific components on the basis of data coming from operating experience gained in various fusion laboratories. The activity began in 2001 with the study of the Joint European Torus vacuum and active gas handling systems. Two years later the

The Safety and Tritium Applied Research (STAR) Facility: Status-2004

International Nuclear Information System (INIS)

Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Sharpe, J.P.; Schuetz, S.T.; Petti, D.A.

2005-01-01

The Safety and Tritium Applied Research (STAR) Facility, a US DOE National User Facility at the Idaho National Engineering and Environmental Laboratory (INEEL), comprises capabilities and infrastructure to support both tritium and non-tritium research activities important to the development of safe and environmentally friendly fusion energy. Research thrusts include (1) interactions of tritium and deuterium with plasma-facing-component (PFC) materials, (2) fusion safety issues [PFC material chemical reactivity and dust/debris generation, activation product mobilization, tritium behavior in fusion systems], and (3) molten salts and fusion liquids for tritium breeder and coolant applications. This paper updates the status of STAR and the capabilities for ongoing research activities, with an emphasis on the development, testing and integration of the infrastructure to support tritium research activities. Key elements of this infrastructure include a tritium storage and assay system, a tritium cleanup system to process glovebox and experiment tritiated effluent gases, and facility tritium monitoring systems

The Testing Behind The Test Facility: The Acoustic Design of the NASA Glenn Research Center's World-Class Reverberant Acoustic Test Facility

Science.gov (United States)

Hozman, Aron D.; Hughes, William O.; McNelis, Mark E.; McNelis, Anne M.

2011-01-01

The National Aeronautics and Space Administration (NASA) Glenn Research Center (GRC) is leading the design and build of the new world-class vibroacoustic test capabilities at the NASA GRC's Plum Brook Station in Sandusky, Ohio, USA. Benham Companies, LLC is currently constructing modal, base-shake sine and reverberant acoustic test facilities to support the future testing needs of NASA's space exploration program. The large Reverberant Acoustic Test Facility (RATF) will be approximately 101,000 cu ft in volume and capable of achieving an empty chamber acoustic overall sound pressure level (OASPL) of 163 dB. This combination of size and acoustic power is unprecedented amongst the world's known active reverberant acoustic test facilities. The key to achieving the expected acoustic test spectra for a range of many NASA space flight environments in the RATF is the knowledge gained from a series of ground acoustic tests. Data was obtained from several NASA-sponsored test programs, including testing performed at the National Research Council of Canada's acoustic test facility in Ottawa, Ontario, Canada, and at the Redstone Technical Test Center acoustic test facility in Huntsville, Alabama, USA. The majority of these tests were performed to characterize the acoustic performance of the modulators (noise generators) and representative horns that would be required to meet the desired spectra, as well as to evaluate possible supplemental gas jet noise sources. The knowledge obtained in each of these test programs enabled the design of the RATF sound generation system to confidently advance to its final acoustic design and subsequent on-going construction.

Tritium experience in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

1998-01-01

Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

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)

New tritium monitor for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Jalbert, R.A.

1985-01-01

At DT-fueled fusion reactors, there will be a need for tritium monitors that can simultaneously measure in real time the concentrations of HTO, HT and the activated air produced by fusion neutrons. Such a monitor has been developed, tested and delivered to the Princeton Plasma Physics Laboratory for use at the Tokamak Fusion Test Reactor (TFTR). It uses semipermeable membranes to achieve the removal of HTO from the sampled air for monitoring and a catalyst to convert the HT to HTO, also for removal and monitoring. The remaining air, devoid of tritium, is routed to a third detector for monitoring the activated air. The sensitivities are those that would be expected from tritium instruments employing conventional flow-through ionization chambers: 1 to 3 μCi/m 3 . Its discriminating ability is approximately 10 -3 for any of the three components (HTO, HT and activated air) in any of the other two channels. For instance, the concentration of HT in the HTO channel is 10 -3 times its original concentration in the sampled air. This will meet the needs of TFTR

Status of the cold test facility for the JT-60SA tokamak toroidal field coils

Energy Technology Data Exchange (ETDEWEB)

Abdel Maksoud, Walid, E-mail: walid.abdelmaksoud@cea.fr; Bargueden, Patrick; Bouty, André; Dispau, Gilles; Donati, André; Eppelle, Dominique; Genini, Laurent; Guiho, Patrice; Guihard, Quentin; Joubert, Jean-Michel; Kuster, Olivier; Médioni, Damien; Molinié, Frédéric; Sinanna, Armand; Solenne, Nicolas; Somson, Sébastien; Vieillard, Laurence

2015-10-15

Highlights: • The 5 K cryogenic loop includes a 500 W refrigerator and a She cold pump. • The coils are energized thanks to a 25.7 kA power supply and HTS current leads. • Temperature margin tests between 5 K and 7.5 K will be made on each coil. • A magnet safety system protects each double pancake of the coil in case of quench. • Instrumentation is monitored on a 1 Hz to 10 kHz fast acquisition system. - Abstract: JT-60SA is a fusion experiment which is jointly constructed by Japan and Europe and which shall contribute to the early realization of fusion energy, by providing support to the operation of ITER, and by addressing key physics issues for ITER and DEMO. In order to achieve these goals, the existing JT-60U experiment will be upgraded to JT-60SA by using superconducting coils. The 18 TF coils of the JT-60SA device will be provided by European industry and tested in a Cold Test Facility (CTF) at CEA Saclay. The coils will be tested at the nominal current of 25.7 kA and will be cooled with supercritical helium between 5 K and 7.5 K to check the temperature margin against a quench. The main objective of these tests is to check the TF coils performance and hence mitigate the fabrication risks. The most important components of the facility are: a 11.5 m × 6.5 m large cryostat in which the TF coils will be thermally insulated by vacuum; a 500 W helium refrigerator and a valve box to cool the coils down to 5 K and circulate 24 g/s of supercritical helium through the winding pack and through the casing; a power supply and HTS current leads to energize the coil; the control and instrumentation equipment (sensors, PLC's, supervision system, fast data acquisition system, etc.) and the Magnet Safety System (MSS) that protects the coils in case of quench. The paper will give an overview of the design of this large facility and the status of its realization.

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

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

Science.gov (United States)

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

2017-11-01

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

Feasibility study of a magnetic fusion production reactor

Science.gov (United States)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells

Tokamak Fusion Test Reactor D-T results

International Nuclear Information System (INIS)

Meade, D.M.

1995-01-01

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, α confinement, α heating and possible α-driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of about 20MW of tritium and 14MW of deuterium neutral beams into the TFTR produced a plasma with a T-to-D density ratio of about 1 and yielding a maximum fusion power of about 9.2MW. The fusion power density in the core of the plasma was about 1.8MWm -3 , approximating that expected in a D-T fusion reactor. A TFTR plasma with a T-to-D density ratio of about 1 was found to have about 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass A of τ E ∝A 0.6 . The core ion temperature increased from 30 to 37keV owing to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 to 10.6keV can be attributed to electron heating by the α particles. The approximately 5% loss of α particles, as observed on detectors near the bottom edge of the plasma, was consistent with classical first orbit loss without anomalous effects. Initial measurements have been made of the confined high energy α particles and the resultant α ash density. At fusion power levels of 7.5MW, fluctuations at the toroidal Alfven eigen-mode frequency were observed by the fluctuation diagnostics. However, no additional α loss due to the fluctuations was observed. (orig.)

Overview of the main challenges for the engineering design of the test facilities system of IFMIF

International Nuclear Information System (INIS)

Molla, J.; Nakamura, K.

2009-01-01

High intense radiation fields were demanded to IFMIF to address the lack of information on effects in materials due to radiation fields with fusion reactor features. Such intense radiation fields will also produce a number of unwanted effects in exposed materials and components. The main difficulties to achieve a reliable engineering design of the Test Facilities System during the Engineering Validation and the Engineering Design phase of IFMIF now under development are reviewed in this paper. The most challenging activities will be the design of the high flux test module, the creep fatigue test module, the test cell and the remote handling system. The intense radiation fields in the irradiation area and the high availability required for IFMIF (70%) are the main reasons for these difficulties.

Importance of tests in nuclear facilities

International Nuclear Information System (INIS)

Guillemard, B.

1985-10-01

In nuclear facilities, safety related systems and equipments are subject, along their whole service-life, to numerous tests. This paper analyses the role of tests in the successive stages of design, construction, exploitation of a nuclear facility. It examines several aspects of test quality control: definition of needs, test planning, intrinsic quality of each test, control of interfaces (test are both the end and the starting point of many actions concerned by quality) and the application [fr

40 CFR 792.43 - Test system care facilities.

Science.gov (United States)

2010-07-01

.... (a) A testing facility shall have a sufficient number of animal rooms or other test system areas, as... different tests. (b) A testing facility shall have a number of animal rooms or other test system areas... waste and refuse or for safe sanitary storage of waste before removal from the testing facility...

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.

Multidirectional testing of one- and two-level ProDisc-L versus simulated fusions.

Science.gov (United States)

Panjabi, Manohar; Henderson, Gweneth; Abjornson, Celeste; Yue, James

2007-05-20

An in vitro human cadaveric biomechanical study. To evaluate intervertebral rotation changes due to lumbar ProDisc-L compared with simulated fusion, using follower load and multidirectional testing. Artificial discs, as opposed to the fusions, are thought to decrease the long-term accelerated degeneration at adjacent levels. A biomechanical assessment can be helpful, as the long-term clinical evaluation is impractical. Six fresh human cadaveric lumbar specimens (T12-S1) underwent multidirectional testing in flexion-extension, bilateral lateral bending, and bilateral torsion using the Hybrid test method. First, intact specimen total range of rotation (T12-S1) was determined. Second, using pure moments again, this range of rotation was achieved in each of the 5 constructs: A) ProDisc-L at L5-S1; B) fusion at L5-S1; C) ProDisc-L at L4-L5 and fusion at L5-S1; D) ProDisc-L at L4-L5 and L5-S1; and E) 2-level fusion at L4-L5 to L5-S1. Significant changes in the intervertebral rotations due to each construct were determined at the operated and nonoperated levels using repeated measures single factor ANOVA and Bonferroni statistical tests (P < 0.05). Adjacent-level effects (ALEs) were defined as the percentage changes in intervertebral rotations at the nonoperated levels due to the constructs. One- and 2-level ProDisc-L constructs showed only small ALE in any of the 3 rotations. In contrast, 1- and 2-level fusions showed increased ALE in all 3 directions (average, 7.8% and 35.3%, respectively, for 1 and 2 levels). In the disc plus fusion combination (construct C), the ALEs were similar to the 1-level fusion alone. In general, ProDisc-L preserved physiologic motions at all spinal levels, while the fusion simulations resulted in significant ALE.

Arc Heated Scramjet Test Facility

Data.gov (United States)

Federal Laboratory Consortium — The Arc Heated Scramjet Test Facility is an arc heated facility which simulates the true enthalpy of flight over the Mach number range of about 4.7 to 8 for free-jet...

Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*

International Nuclear Information System (INIS)

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Jaeger, E.F.; Janos, A.; Jassby, D.L.; Jobes, F.C.; Johnson, D.W.; Johnson, L.C.; Kamperschroer, J.; Kesner, J.; Kugel, H.; Kwon, S.; Labik, G.; Lam, N.T.; LaMarche, P.H.; Laughlin, M.J.; Lawson, E.; LeBlanc, B.; Leonard, M.; Levine, J.; Levinton, F.M.; Loesser, D.; Long, D.; Machuzak, J.; Mansfield, D.E.; Marchlik, M.; Marmar, E.S.; Marsala, R.; Martin, A.; Martin, G.; Mastrocola, V.; Mazzucato, E.; McCarthy, M.P.; Majeski, R.; Mauel, M.; McCormack, B.; McCune, D.C.; McGuire, K.M.; Meade, D.M.; Medley, S.S.; Mikkelsen, D.R.; Milora, S.L.; Monticello, D.; Mueller, D.; Murakami, M.; Murphy, J.A.; Nagy, A.; Navratil, G.A.; Nazikian, R.; Newman, R.; Nishitani, T.; Norris, M.; O'Connor, T.; Oldaker, M.; Ongena, J.; Osakabe, M.; Owens, D.K.; Park, H.; Park, W.; Paul, S.F.; Pavlov, Y.I.; Pearson, G.; Perkins, F.; Perry, E.; Persing, R.; Petrov, M.; Phillips, C.K.; Pitcher, S.; Popovichev, S.; Qualls, A.L.; Raftopoulos, S.; Ramakrishnan, R.; Ramsey, A.; Rasmussen, D.A.; Redi, M.H.

1994-01-01

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert TM system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of ∼10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of α-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined α particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed

40 CFR 160.43 - Test system care facilities.

Science.gov (United States)

2010-07-01

... testing facility shall have a number of animal rooms or other test system areas separate from those... housed, facilities shall exist for the collection and disposal of all animal waste and refuse or for safe sanitary storage of waste before removal from the testing facility. Disposal facilities shall be so...

Drop test facility available to private industry

International Nuclear Information System (INIS)

Shappert, L.B.; Box, W.D.

1983-01-01

In 1978, a virtually unyielding drop test impact pad was constructed at Oak Ridge National Laboratory's (ORNL's) Tower Shielding Facility (TSF) for the testing of heavy shipping containers designed for transporting radioactive materials. Because of the facility's unique capability for drop-testing large, massive shipping packages, it has been identified as a facility which can be made available for non-DOE users

Tokamak fusion test reactor FELIX plate experiment

International Nuclear Information System (INIS)

Hua, T.O.; Nygren, R.E.; Turner, L.R.

1986-01-01

For a conducting material exposed to both a time-varying and a static magnetic field, such as a limiter blade in a tokamak, the induced eddy currents and the deflection arising from those eddy currents can be strongly coupled. The coupling effects reduce the currents and deflections markedly, sometimes an order of magnitude, from the values predicted if coupling is neglected. A series of experiments to study current-deflection coupling were performed using the Fusion Electromagnetic Inductance Experiment (FELIX) facility at Argonne National Laboratory. Magnetic damping and magnetic stiffness resulting from the coupling are discussed, and analytical expressions for induced eddy current and rigid body rotation in the FELIX plate experiment are compared with the experimental results. Predictions for the degree of coupling based on various parameters are made using the analytical model

Plasma Facing Components Generic Facilities Review Panel (PFC-GFRP): Final report

International Nuclear Information System (INIS)

McGrath, R.; Allen, S.; Hill, D.; Brooks, J.; Mattas, R.; Davis, J.; Lipschultz, B.; Ulrickson, M.

1993-10-01

The Plasma Facing Components (PFC) Facilities Review Panel was chartered by the US Department of Energy, Office of Fusion Energy, ITER (International Thermonuclear Experimental Reactor) and Technology Division, to outline the program plan and identify the supporting test facilities that lead to reliable, long-lived plasma facing components for ITER. This report summarizes the panel's findings and identifies the necessary and sufficient set of test facilities required for ITER PFC development

Fusion power: massive research program aims at formidable problems, almost limitless potential

International Nuclear Information System (INIS)

Dingee, D.A.

1979-01-01

This article surveys extensively fusion development under the following topics: US research directions; inertial confinement fusion; foreign fusion efforts; fusion issues; fusion applications; and arguments for fusion development. Dr. Dingee points out that, despite persuasive arguments for development, fusion has as yet attracted no substantial constituency; and that winning greater support for fusion may thus require a considerable technical breakthrough (namely, proof of scientific feasibility or achievement of energy breakeven) - or a new focus on an energy source such as hybrids, which offer a nearer-term payoff than pure fusion. Dr. Dingee says the next major facility for magnetic confinement research (to be built in late 1980s) has not yet been selected, but will probably be an engineering test facility; there are similar plans for inertial confinement. Whichever type is chosen, the first experimental power reactor is scheduled for the first few years of the 2000's, this to be followed by commercial demonstration of fusion power in the 2010 to 2020 time frame. He points out, finally, that the complex technical and institutional issues are being considered in a climate in which the benefits of nuclear energy itself are being questioned; and that there is little doubt that future development is tied to overall decisions the nation will make regarding the value of nuclear energy

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.

Toroid magnet test facility

CERN Multimedia

2002-01-01

Because of its exceptional size, it was not feasible to assemble and test the Barrel Toroid - made of eight coils - as an integrated toroid on the surface, prior to its final installation underground in LHC interaction point 1. It was therefore decided to test these eight coils individually in a dedicated test facility.

STG-ET: DLR electric propulsion test facility

Directory of Open Access Journals (Sweden)

Andreas Neumann

2017-04-01

Full Text Available DLR operates the High Vacuum Plume Test Facility Göttingen – Electric Thrusters (STG-ET. This electric propulsion test facility has now accumulated several years of EP-thruster testing experience. Special features tailored to electric space propulsion testing like a large vacuum chamber mounted on a low vibration foundation, a beam dump target with low sputtering, and a performant pumping system characterize this facility. The vacuum chamber is 12.2m long and has a diameter of 5m. With respect to accurate thruster testing, the design focus is on accurate thrust measurement, plume diagnostics, and plume interaction with spacecraft components. Electric propulsion thrusters have to run for thousands of hours, and with this the facility is prepared for long-term experiments. This paper gives an overview of the facility, and shows some details of the vacuum chamber, pumping system, diagnostics, and experiences with these components.

Design of electronic measurement and quench detection equipment for the Current Lead Test facility Karlsruhe (CuLTKa)

International Nuclear Information System (INIS)

Hollik, Markus; Fietz, Walter H.; Fink, Stefan; Gehrlein, Mirko; Heller, Reinhard; Lange, Christian; Möhring, Tobias

2013-01-01

The Current Lead Test facility Karlsruhe (CuLTKa) is under construction at the Karlsruhe Institute of Technology (KIT) to perform acceptance tests of high temperature superconductor (HTS) current leads (CL). CuLTKa is in progress and present planning expects the completion in 2013. The data acquisition system is based on a modular design with electronic measurement and monitoring equipment covering a test voltage of 50 kV DC against ground. It provides plug-in units which enable temperature and voltage measurement at high voltage potential and in addition quench detection units which detect a loss of superconductivity reliably and quickly to avoid damage of the superconducting device under test. Prototype units for quench detection, temperature and voltage measurement have been successfully tested. Six temperature measurement units are already in use in the KIT test facility TOSKA and operated reliably during the acceptance tests of the HTS current leads for Wendelstein 7-X (W7-X) in 2011/2012. CuLTKa will be used first for 26 current leads which will be built in KIT for the fusion experiment JT-60SA. The present paper gives an overview of the design of the electronic measurement and quench detection equipment

Thermal-hydraulic tests with out-of-pile test facility for BOCA development

International Nuclear Information System (INIS)

Kitagishi, Shigeru; Aoyama, Masashi; Tobita, Masahiro; Inaba, Yoshitomo; Yamaura, Takayuki

2012-01-01

The fuel transient test facility was prepared for power ramping tests of light-water-reactor (LWR) fuels in the Japan Materials Testing Reactor (JMTR) under a contract project with the Nuclear Industrial Safety Agent (NISA) of the Ministry of Economy, Trade and Industry (METI). It is necessary to develop high accuracy analysis procedure for power ramping tests after restart of the JMTR. The out-of-pile test facility to simulate thermal-hydraulic conditions of the fuel transient test facility was therefore developed. Applicability of the analysis code ACE-3D was examined for thermal-hydraulic analysis of power ramping tests for 10x10 BWR fuels by the fuel transient test facility. As the results, the calculated temperature was 304°C in comparison with measured value of 304.9-317.4°C in the condition of 600 W/cm. There is a bright prospect of high accuracy power ramping tests by the fuel transient test facility in JMTR. (author)

Energy sweepstakes: fusion gets a chance

International Nuclear Information System (INIS)

Robinson, A.L.

1980-01-01

Congress plans to speed up the magnetic-fusion program by shifting the emphasis from plasma research to fusion-reactor engineering. The bill doubles the overall fusion budget over the next five years in order to construct a Fusion Engineering Device (FED) by 1990. A review panel of scientists suggested limiting the cost to under $1 billion and holding the increase until late 1983. The panel also suggested waiting until 1990 to set a date for demonstrating a competitive commercial reactor even though progress made in the 1970s could bring a realistic date as close as 2000. The new policy evolves from the debate between tokamak hawks, who want to take the best prospect to commercialization immediately, and the doves, who want to wait to see if the best possible concept turns out to be the magnetic mirror or some other contender. The Engineering Test Facility (ETF) represents a compromise of these positions

Comparative assessment of world research efforts on magnetic confinement fusion

International Nuclear Information System (INIS)

McKenney, B.L.; McGrain, M.; Rutherford, P.H.

1990-02-01

This report presents a comparative assessment of the world's four major research efforts on magnetic confinement fusion, including a comparison of the capabilities in the Soviet Union, the European Community (Western Europe), Japan, and the United States. A comparative evaluation is provided in six areas: tokamak confinement; alternate confinement approaches; plasma technology and engineering; and fusion computations. The panel members are involved actively in fusion-related research, and have extensive experience in previous assessments and reviews of the world's four major fusion programs. Although the world's four major fusion efforts are roughly comparable in overall capabilities, two conclusions of this report are inescapable. First, the Soviet fusion effort is presently the weakest of the four programs in most areas of the assessment. Second, if present trends continue, the United States, once unambiguously the world leader in fusion research, will soon lose its position of leadership to the West European and Japanese fusion programs. Indeed, before the middle 1990s, the upgraded large-tokamak facilities, JT-60U (Japan) and JET (Western Europe), are likely to explore plasma conditions and operating regimes well beyond the capabilities of the TFTR tokamak (United States). In addition, if present trends continue in the areas of fusion nuclear technology and materials, and plasma technology and materials, and plasma technology development, the capabilities of Japan and Western Europe in these areas (both with regard to test facilities and fusion-specific industrial capabilities) will surpass those of the United States by a substantial margin before the middle 1990s

Mirror Fusion vacuum technology developments

International Nuclear Information System (INIS)

Batzer, T.H.; Call, W.R.

1983-01-01

Magnetic Mirror Fusion experiments, such as MFTF-B+T (Mirror Fusion Test Facility-B, Tritium Upgrade) and foreseeable follow-on devices, have operational and maintenance requirements that have not yet been fully demonstrated. Among those associated with vacuum technology are the very-high continuous-pumping speeds, 10 7 to 10 8 l/s for D 2 , T 2 and, to a lesser extent, He; the early detection of water leaks from the very-high heat-flux neutral-beam dumps and the detection and location of leaks in the superconducting magnets not protected by guard vacuums. Possible solutions to these problems have been identified and considerable progress has been made toward successfully demonstrating their feasibility