Fusion reactor design studies

International Nuclear Information System (INIS)

Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

1990-01-01

This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources

Fusion safety data base

International Nuclear Information System (INIS)

Laats, E.T.; Hardy, H.A.

1983-01-01

The purpose of this Fusion Safety Data Base Program is to provide a repository of data for the design and development of safe commercial fusion reactors. The program is sponsored by the United States Department of Energy (DOE), Office of Fusion Energy. The function of the program is to collect, examine, permanently store, and make available the safety data to the entire US magnetic-fusion energy community. The sources of data will include domestic and foreign fusion reactor safety-related research programs. Any participant in the DOE Program may use the Data Base Program from his terminal through user friendly dialog and can view the contents in the form of text, tables, graphs, or system diagrams

Safety methodology implementation in the conceptual design phase of a fusion reactor

International Nuclear Information System (INIS)

Rodriguez-Rodrigo, L.; Elbez-Uzan, J.

2007-01-01

The licensing of ITER in France represents the first process for licensing a fusion facility in the framework of an experimental device with a total Tritium inventory of 3 kg. The main ITER parameters are far from those expected in the future demonstration reactors where the fusion power will be at least 5 times higher and the additional heating power could also reach up to 5 times the one foreseen in ITER. Main safety requirements for these reactors are based, among other conditions, on their inherent features as low amount of fuel, very low impurity content of structural materials, minimum waste repository, no active systems for safe shut-down, and no need for evacuation of population after the most severe accident. The design of such reactors is at the stage of conceptual studies and is mainly dealing with plasma performances, tritium breeding, blanket/divertor designs and solution of engineering issues, as well as bounding accidents or classification of waste. The methodological approach for integrating safety analysis as a tool for optimizing the design of the overall fusion installation for future reactors in the conceptual design phase is sketched, including the machine itself and the different auxiliary nuclear buildings. (author)

Tritium management in fusion reactors

International Nuclear Information System (INIS)

Galloway, T.R.

1978-05-01

This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

Safety considerations of lithium lead alloy as a fusion reactor breeding material

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.

1985-01-01

Test results and conclusions are presented for lithium lead alloy interactions with various gas atmospheres, concrete and potential reactor coolants. The reactions are characterized to evaluate the potential of volatilizing and transporting radioactive species associated with the liquid breeder under postulated fusion reactor accident conditions. The safety concerns identified for lithium lead alloy reactions with the above materials are compared to those previously identified for a reference fusion breeder material, liquid lithium. Conclusions made from this comparison are also included

Thermal safety analysis for pebble bed blanket fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie

1998-01-01

Pebble bed blanket hybrid reactor may have more advantages than slab element blanket hybrid reactor in nuclear fuel production and nuclear safety. The thermo-hydraulic calculations of the blanket in the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor developed in China are carried out using the Code THERMIX and auxiliary code. In the calculations different fuel pebble material and steady state, depressurization and total loss of flow accident conditions are included. The results demonstrate that the conceptual design of the Tokamak helium cooling pebble bed blanket fusion-fission hybrid reactor with dump tank is feasible and safe enough only if the suitable fuel pebble material is selected and the suitable control system and protection system are established. Some recommendations for due conceptual design are also presented

Discussion on Safety Analysis and Regulatory Framework for the Future Fusion Reactors

Energy Technology Data Exchange (ETDEWEB)

Kang, Myoung-suk; Oh, Kyemin; Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of)

2015-05-15

This study aims to secure the core original technologies and expand the base of domestic specialist at a fusion area by pursuing and developing nonprocurement technologies for ITER. From this project, the latest technical data and experiences have been recorded for the development of the safety regulation and safety-related design criteria of the future fusion reactors in Korea. In this context, this paper discusses on the progress of surveying the ITER licensing process and regulatory issues revealed. The regulation and licensing process for a fusion power plant has been expected to be quite different due to unique and unforeseen properties differently from the conventional nuclear facilities. To overcome this, not only various safety issues should be analyzed, but safety objectives, regulatory requirements, and design variables should also be established in detailed design phase. We expect our survey will contribute on the discussion to establish general and technical safety principles for national fusion power plant technology plans.

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

On fusion and fission breeder reactors

International Nuclear Information System (INIS)

Brandt, B.; Schuurman, W.; Klippel, H.Th.

1981-02-01

Fast breeder reactors and fusion reactors are suitable candidates for centralized, long-term energy production, their fuel reserves being practically unlimited. The technology of a durable and economical fusion reactor is still to be developed. Such a development parallel with the fast breeder is valuable by reasons of safety, proliferation, new fuel reserves, and by the very broad potential of the development of the fusion reactor. In order to facilitate a discussion of these aspects, the fusion reactor and the fast breeder reactor were compared in the IIASA-report. Aspects of both reactor systems are compared

Fusion reactors as a future energy source

International Nuclear Information System (INIS)

Seifritz, W.

A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors, (8) fuel supply for fusion reactors, (9) safety and environmental impact, and (10) accidents, and (11) waste removal and storage

Lower activation materials and magnetic fusion reactors

International Nuclear Information System (INIS)

Conn, R.W.; Bloom, E.E.; Davis, J.W.; Gold, R.E.; Little, R.; Schultz, K.R.; Smith, D.L.; Wiffen, F.W.

1984-01-01

Radioactivity in fusion reactors can be effectively controlled by materials selection. The detailed relationship between the use of a material for construction of a magnetic fusion reactor and the material's characteristics important to waste disposal, safety, and system maintainability has been studied. The quantitative levels of radioactivation are presented for many materials and alloys, including the role of impurities, and for various design alternatives. A major outcome has been the development of quantitative definitions to characterize materials based on their radioactivation properties. Another key result is a four-level classification scheme to categorize fusion reactors based on quantitative criteria for waste management, system maintenance, and safety. A recommended minimum goal for fusion reactor development is a reference reactor that (a) meets the requirements for Class C shallow land burial of waste materials, (b) permits limited hands-on maintenance outside the magnet's shield within 2 days of a shutdown, and (c) meets all requirements for engineered safety. The achievement of a fusion reactor with at least the characteristics of the reference reactor is a realistic goal. Therefore, in making design choices or in developing particular materials or alloys for fusion reactor applications, consideration must be given to both the activation characteristics of a material and its engineering practicality for a given application

The LOFA analysis of fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Yu, Z.-C.; Xie, H.

2014-01-01

The fusion-fission hybrid energy reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc, with the fusion neutron source striking the subcritical blanket. The passive safety system, consisting of passive residual heat removal system, passive safety injection system and automatic depressurization system, was adopted into the fusion-fission hybrid energy reactor in this paper. Modeling and nodalization of primary loop, passive core cooling system and partial secondary loop of the fusion-fission hybrid energy reactor using RELAP5 were conducted and LOFA (Loss of Flow Accident) was analyzed. The results of key transient parameters indicated that the PRHRs could mitigate the accidental consequence of LOFA effectively. It is also concluded that it is feasible to apply the passive safety system concept to fusion-fission hybrid energy reactor. (author)

Fusion safety program plan

International Nuclear Information System (INIS)

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

1980-09-01

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

Fusion Safety Program annual report, Fiscal Year 1993

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.

1993-12-01

This report summarizes the major activities of the Fusion Safety Program in Fiscal Year 1993. The Idaho National Engineering Laboratory (INEL) has been designated by DOE as the lead laboratory for fusion safety, and EG ampersand G Idaho, Inc., is the prime contractor for INEL operations. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL and in participating organizations, including universities and private companies. Technical areas covered in the report include tritium safety, beryllium safety, activation product release, reactions involving potential plasma-facing materials, safety of fusion magnet systems, plasma disruptions and edge physics modeling, risk assessment failure rates, computer codes for reactor transient analysis, and regulatory support. These areas include work completed in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed at the INEL for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor projects at the Princeton Plasma Physics Laboratory and a summary of the technical support for the ARIES/PULSAR commercial reactor design studies

Flibe Use in Fusion Reactors - An Initial Safety Assessment

Energy Technology Data Exchange (ETDEWEB)

Cadwallader, Lee Charles; Longhurst, Glen Reed

1999-04-01

This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF2) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

Flibe use in fusion reactors: An initial safety assessment

International Nuclear Information System (INIS)

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

1999-01-01

This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF 2 ) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material

Assessment of fusion reactor development. Proceedings

International Nuclear Information System (INIS)

Inoue, N.; Tazima, T.

1994-04-01

Symposium on assessment of fusion reactor development was held to make clear critical issues, which should be resolved for the commercial fusion reactor as a major energy source in the next century. Discussing items were as follows. (1) The motive force of fusion power development from viewpoints of future energy demand, energy resources and earth environment for 'Sustainable Development'. (2) Comparison of characteristics with other alternative energy sources, i.e. fission power and solar cell power. (3) Future planning of fusion research and advanced fuel fusion (D 3 He). (4) Critical issues of fusion reactor development such as Li extraction from the sea water, structural material and safety. (author)

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

Flibe use in fusion reactors -- An initial safety assessment

Energy Technology Data Exchange (ETDEWEB)

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

1999-03-01

This report is an initial effort to identify and evaluate safety issues associated with the use of Flibe (LiF-BeF{sub 2}) as a molten salt coolant for nuclear fusion power plant applications. Flibe experience in the Molten Salt Reactor Experiment is briefly reviewed. Safety issues identified include chemical toxicity, radiological issues resulting from neutron activation, and the operational concerns of handling a high temperature coolant. Beryllium compounds and fluorine pose be toxicological concerns. Some controls to protect workers are discussed. Since Flibe has been handled safely in other applications, its hazards appear to be manageable. Some safety issues that require further study are pointed out. Flibe salt interaction with strong magnetic fields should be investigated. Evolution of Flibe constituents and activation products at high temperature (i.e., will Fluorine release as a gas or remain in the molten salt) is an issue. Aerosol and tritium release from a Flibe spill requires study, as does neutronics analysis to characterize radiological doses. Tritium migration from Flibe into the cooling system is also a safety concern. Investigation of these issues will help determine the extent to which Flibe shows promise as a fusion power plant coolant or plasma-facing material.

Fusion Safety Program annual report, fiscal year 1992

International Nuclear Information System (INIS)

Holland, D.F.; Cadwallader, L.C.; Herring, J.S.; Longhurst, G.R.; McCarthy, K.A.; Merrill, B.J.; Piet, S.J.

1993-01-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1992. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and EG ampersand G Idaho, Inc. is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL and in participating organizations including the Westinghouse Hanford Company at the Hanford Engineering Development Laboratory, the Massachusetts Institute of Technology, and the University of Wisconsin. The technical areas covered in the report include tritium safety, activation product release, reactions involving beryllium, reactions involving lithium breeding materials, safety of fusion magnet systems, plasma disruptions, risk assessment failure rate data base, and computer code development for reactor transients. Also included in the report is a summary of the safety and environmental studies performed by the INEL for the Tokamak Physics Experiments and the Tokamak Fusion Test Reactor, the safety analysis for the International Thermonuclear Experimental Reactor design, and the technical support for the ARIES commercial reactor design study

Environmental aspects of fusion reactors

International Nuclear Information System (INIS)

Coffman, F.E.; Williams, J.M.

1975-01-01

With the continued depletion of fossil and uranium resources in the coming decades, the U. S. will be forced to look more toward renewable energy resources (e.g., wind, tidal, geothermal, and solar power) and toward such longer-term and nondepletable energy resources as fissile fast breeder reactors and fusion power. Several reference reactor designs have been completed for full-scale fusion power reactors that indicate that the environmental impacts from construction, operation, and eventual decommissioning of fusion reactors will be quite small. The principal environmental impact from fusion reactor operation will be from thermal discharges. Some of the safety and environmental characteristics that make fusion reactors appear attractive include an effectively infinite fuel supply at low cost, inherent incapability for a ''nuclear explosion'' or a ''nuclear runaway,'' the absence of fission products, the flexibility of selecting low neutron-cross-section structural materials so that emergency core cooling for a loss-of-coolant or other accident will not be necesary, and the absence of special nuclear materials such as 235 U or 239 Pu, so that diversion of nuclear weapons materials will not be possible and nuclear blackmail will not be a serious concern

Implications of recent implantation-driven permeation experiments for fusion reactor safety

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Struttmann, D.A.

1986-01-01

Metal structures exposed to the plasma in tritium-burning fusion reactors will be subject to implantation-driven permeation (IDP) of tritium. Permeation rates for IDP in fusion structural materials are usually high because the tritium atoms enter the material without having to go through the dissociation and solution steps required of tritium-bearing gas molecules. These surface processes, which may be rate limiting in PDP, actually enhance permeation in IDP by inhibiting the return of tritium to the plasma side of the structure. Experiments have been conducted at the Idaho National Engineering Laboratory (INEL) to investigate the nature of IDP by simulating conditions experienced by structures exposed to the plasma. These experiments have shown that surface conditions are important to tritium permeation in materials endothermic to hydrogen solution such as austenitic and ferritic steels. In reactive metals such as vanadium, surface processes appear to totally control the permeation. The purpose of this paper is to review the progress of those experiments and to discuss the implications that the results have regarding the tritium-related safety concerns of fusion reactors

Safety analysis on tokamak helium cooling slab fuel fusion-fission hybrid reactor

International Nuclear Information System (INIS)

Wei Renjie; Jian Hongbing

1992-01-01

The thermal analyses for steady state, depressurization and total loss of flow in the tokamak helium cooling slab fuel element fusion-fission hybrid reactor are presented. The design parameters, computed results of HYBRID program and safety evaluation for conception design are given. After all, it gives some recommendations for developing the design

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

Fusion reactors and the environment

International Nuclear Information System (INIS)

Wrixon, A.D.

1976-01-01

A summary is given of the report of a study group set up in 1971 by the Director of the UKAEA Culham Laboratory to investigate environmental and safety aspects of future commercial fusion reactors (1975, Carruthers, R., Dunster, H.J., Smith, R.D., Watson, C.J.H., and Mitchell, J.T.D., Culham Study Group Report on Fusion Reactors and the Environment, CLM-R148, HMSO, London). This report was originally issued in 1973 under limited distribution, but has only recently been made available for open circulation. Deuterium/tritium fusion is thought to be the most likely reaction to be used in the first generation of reactors. Estimates were made of the local and world-wide population hazards from the release of tritium, both under normal operating conditions and in the event of an accident. One serious type of accident would be a lithium metal fire in the blanket region of the reactor. The use of a fusible lithium salt (FLIBE), eliminating the lithium fire risk, is considered but the report concentrates on lithium metal in the blanket region. The main hazards to operating staff arise both from tritium and from neutron activation of the construction materials. Remote servicing of the reactor structure will be essential, but radioactive waste management seems less onerous than for fission reactors. Meaningful comparison of the overall hazards associated with fusion and fission power programmes is not yet possible. The study group emphasized the need for more data to aid the safety assessments, and the need for such assessments to keep pace with fusion power station design. (U.K.)

Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

International Nuclear Information System (INIS)

Ho, S.K.; Fowler, T.K.; Holdren, J.P.

1991-11-01

This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications

Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

Safety analysis and environmental effects of fusion concepts

International Nuclear Information System (INIS)

Anon.

1976-01-01

Fusion reactor concepts have been analyzed to determine the probable interactions with the environment and the resultant environmental effects. Two research projects on tritium oxidation in the atmosphere and carbon-14 formation in fusion reactors are briefly described. A study and report were completed, investigating the potential public safety impact of accidents in fusion power plants. After reviewing the existing information on conceptual fusion reactor designs, PNL identified areas of safety concern, making recommendations on how development of safety information might be best accomplished. Inventories of potentially dispersible toxic materials were classified, and general conclusions were made about their relative importance. The report specifies energy sources with a potential to initiate or propagate an accident. An important product of the study was an assessment logic developed to identify potential accident scenarios that could lead to the release of contaminants to the environment. Though the limited amount of fusion design information allows only a general assessment of accident-initiating events, the logic provides a method for making more detailed safety analyses as more design information becomes available. The same logic was used to identify technological areas where an R and D investment would enhance the technical bases for fusion designs as well as the understanding of safety implications in fusion systems

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.

Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 92--94). Final report

International Nuclear Information System (INIS)

Ho, S.K.; Fowler, T.K.; Holdren, J.P.

1994-01-01

This is the Final Report for a three-year (FY 92--94) study of the Environmental, Safety, and Economic (ESE) aspects of fusion energy systems, emphasizing development of computerized approaches suitable for incorporation as modules in fusion system design codes. First, as is reported in Section 2, the authors now have operating a simplified but complete environment and safety evaluation code, BESAFE. The first tests of BESAFE as a module of the SUPERCODE, a design optimization systems code at LLNL, are reported in Section 3. Secondly, as reported in Section 4, the authors have maintained a strong effort in developing fast calculational schemes for activation inventory evaluation. In addition to these major accomplishments, considerable progress has been made on research on specific topics as follows. A tritium modeling code TRIDYN was developed in collaboration with the TSTA group at LANL and the Fusion Nuclear Technology group at UCLA. A simplified algorithm has been derived to calculate the transient temperature profiles in the blanket during accidents. The scheme solves iteratively a system of non-linear ordinary differential equations describing about 10 regions of the blanket by preserving energy balance. The authors have studied the physics and engineering aspects of divertor modeling for safety applications. Several modifications in the automation and characterization of environmental and safety indices have been made. They have applied this work to the environmental and safety comparisons of stainless steel with alternative structural materials for fusion reactors. A methodology in decision analysis utilizing influence and decision diagrams has been developed to model fusion reactor design problems. Most of the work during this funding period has been reported in 26 publications including theses, journal publications, conference papers, and technical reports, as listed in Section 11

Progress in the development of methodology for fusion safety systems studies

International Nuclear Information System (INIS)

Ho, S.K.; Cambi, G.; Ciattaglia, S.; Fujii-e, Y.; Seki, Y.

1994-01-01

The development of fusion safety systems-study methodology, including the aspects of schematic classification of overall fusion safety system, qualitative assessment of fusion system for identification of critical accident scenarios, quantitative analysis of accident consequences and risk for safety design evaluation, and system-level analysis of accident consequences and risk for design optimization, by a consortium of international efforts is presented. The potential application of this methodology into reactor design studies will facilitate the systematic assessment of safety performance of reactor designs and enhance the impacts of safety considerations on the selection of design configurations

Safety considerations in next step fusion design and beyond

International Nuclear Information System (INIS)

Holland, D.F.

1990-01-01

Recent U.S. and international design studies provide insights into the potential safety and environmental advantages of fusion as well as the development needed to realize this potential. We in the Fusion Safety Program at EG ampersand G Idaho have analyzed the Compact Ignition Tokamak (CIT), the International Thermonuclear Engineering Reactor (ITER), and the Advanced Reactor Innovative Engineering Study (ARIES). I have reviewed these three designs to determine issues related to meeting the safety and the environmental goals that guide fusion development in the U.S. The paper lists safety and environmental issues that are generic to fusion and approaches to favorably resolve each issue. The technical developments that have the highest potential of contributing to improving the safety and environmental attractiveness of fusion are identified and discussed. These developments are in the areas of low-activation materials, plasma- facing components, and plasma physics relating to off-normal plasma events and tritium burn-up. 8 refs., 7 tabs

Canadian contributions to the safety and environmental aspects of fusion

International Nuclear Information System (INIS)

Stasko, R.; Wong, K.

1987-05-01

Since next-step fusion devices will be fuelled with mixtures of tritium and deuterium, the knowledge base and tritium handling experience associated with the operation of CANDU reactors is viewed as relevant to the development of safe fusion technology. Fusion safety issues will be compared with fission safety experience, after which specific Canadian activities in support of fusion safety will be overviewed. In addition, recommendations for appropriate fusion safety criteria will be summarized. 18 refs

Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

International Nuclear Information System (INIS)

Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

1984-09-01

The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits

Conceptual design of a hybrid fusion-fission reactor with intrinsic safety and optimized energy productivity

International Nuclear Information System (INIS)

Talebi, Hosein; Sadat Kiai, S.M.

2017-01-01

Highlights: • Designing a high yield and feasible Dense Plasma Focus for driving the reactor. • Presenting a structural method to design the dual layer cylindrical blankets. • Finding, the blanket production energy, in terms of its geometrical and material parameters. • Designing a subcritical blanket with optimization of energy amplification in detail. - Abstract: A hybrid fission-fusion reactor with a Dense Plasma Focus (DPF) as a fusion core and the dual layer fissionable blanket as the energy multiplier were conceptually designed. A cylindrical DPF, energized by a 200 kJ bank energy, is considered to produce fusion neutron, and these neutrons drive the subcritical fission in the surrounding blankets. The emphasis has been placed on the safety and energy production with considering technical and economical limitations. Therefore, the k eff-t of the dual cylindrical blanket was defined and mathematically, specified. By applying the safety criterion (k eff-t ≤ 0.95), the geometrical and material parameters of the blanket optimizing the energy amplification were obtained. Finally, MCNPX code has been used to determine the detailed dimensions of the blankets and fuel rods.

Fusion Reactor Safety Research Program annual report, FY-79

International Nuclear Information System (INIS)

Crocker, J.G.; Cohen, S.

1980-08-01

The objective of the program is the development, coordination, and execution of activities related to magnetic fusion devices and reactors that will: (a) identify and evaluate potential hazards, (b) assess and disclose potential environmental impacts, and (c) develop design standards and criteria that eliminate, mitigate, or reduce those hazards and impacts. The program will provide a sound basis for licensing fusion reactors. Included in this report are portions of four reports from two outside contractors, discussions of the several areas in which EG and G Idaho is conducting research activities, a discussion of proposed program plan development, mention of special tasks, a review of fusion technology program coordination by EG and G with other laboratories, and a brief view of proposed FY-80 activities

Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Honda, T.; Maki, K.; Okazaki, T.

1994-01-01

Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

Tritium production, management and its impact on safety for a D-3He fusion reactor

International Nuclear Information System (INIS)

Sze, D.K.; Herring, S.; Sawan, M.

1991-11-01

About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

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

Multiphase flow problems on thermofluid safety for fusion reactors

International Nuclear Information System (INIS)

Takase, Kazuyuki

2003-01-01

As the thermofluid safety study for the International Thermonuclear Experimental Reactor (ITER), thermal-hydraulic characteristics of Tokamak fusion reactors under transient events were investigated experimentally and analyzed numerically. As severe transient events an ingress-of-coolant event (ICE) and a loss-of-vacuum event (LOVA) were considered. An integrated ICE test facility was constructed to demonstrate that the ITER safety design approach and parameters are adequate. Water-vapor two-phase flow behavior and performance of the ITER pressure suppression system during the ICE were clarified by the integrated ICE experiments. The TRAC was modified to specify the two-phase flow behavior under the ICE. The ICE experimental results were verified using the modified TRAC code. On the other hand, activated dust mobilization and air ingress characteristics in the ITER vacuum vessel during the LOVA were analyzed using a newly developed analysis code. Some physical models on the motion of dust were considered. The rate of dust released from the vacuum vessel through breaches to the outside was characterized quantitatively. The predicted average pressures in the vacuum vessel during the LOVA were in good agreement with the experimental results. Moreover, direct-contact condensation characteristics between water and vapor inside the ITER suppression tank were observed visually and simulated by the direct two-phase flow analysis. Furthermore, chemical reaction characteristics between vapor and ITER plasma-facing component materials were predicted numerically in order to obtain qualitative estimation on generation of inflammable gases such as hydrogen and methane. The experimental and numerical results of the present studies were reflected in the ITER thermofluid safety design. (author)

Designing the Cascade inertial confinement fusion reactor

International Nuclear Information System (INIS)

Pitts, J.H.

1987-01-01

The primary goal in designing inertial confinement fusion (ICF) reactors is to produce electrical power as inexpensively as possible, with minimum activation and without compromising safety. This paper discusses a method for designing the Cascade rotating ceramic-granule-blanket reactor (Pitts, 1985) and its associated power plant (Pitts and Maya, 1985). Although focus is on the cascade reactor, the design method and issues presented are applicable to most other ICF reactors

The TITAN Reversed-Field Pinch fusion reactor study

International Nuclear Information System (INIS)

1988-03-01

The TITAN Reversed-Field Pinch (RFP) fusion reactor study is a multi-institutional research effort to determine the technical feasibility and key developmental issues of an RFP fusion reactor, especially at high power density, and to determine the potential economics, operations, safety, and environmental features of high-mass-power-density fusion systems. The TITAN conceptual designs are DT burning, 1000 MWe power reactors based on the RFP confinement concept. The designs are compact, have a high neutron wall loading of 18 MW/m 2 and a mass power density of 700 kWe/tonne. The inherent characteristics of the RFP confinement concept make fusion reactors with such a high mass power density possible. Two different detailed designs have emerged: the TITAN-I lithium-vanadium design, incorporating the integrated-blanket-coil concept; and the TITAN-II aqueous loop-in-pool design with ferritic steel structure. This report contains a collection of 16 papers on the results of the TITAN study which were presented at the International Symposium on Fusion Nuclear Technology. This collection describes the TITAN research effort, and specifically the TITAN-I and TITAN-II designs, summarizing the major results, the key technical issues, and the central conclusions and recommendations. Overall, the basic conclusions are that high-mass power-density fusion reactors appear to be technically feasible even with neutron wall loadings up to 20 MW/m 2 ; that single-piece maintenance of the FPC is possible and advantageous; that the economics of the reactor is enhanced by its compactness; and the safety and environmental features need not to be sacrificed in high-power-density designs. The fact that two design approaches have emerged, and others may also be possible, in some sense indicates the robustness of the general findings

Fusion Safety Program annual report, fiscal year 1985

International Nuclear Information System (INIS)

Holland, D.F.; Merrill, B.J.; Herring, J.S.; Piet, S.J.; Longhurst, G.R.

1987-02-01

The Fusion Safety Program (FSP) has supported magnetic fusion technology for seven years, and this is the seventh annual report issued by the FSP. Program focus is identification of the magnitude and distribution of radioactive inventories in fusion reactors, and research and analysis of postulated accident scenarios that could cause the release of a portion of these inventories. Research results are used to develop improved designs that can reduce the probability and magnitude of such releases and thus improve the overall safety of fusion reactors. During FY-1985, research activities continued and participation continued on the Ignition Systems Project (ISP). This report presents the significant results of EGandG Idaho, Inc., activities and those from outside contracts, and includes a list of publications produced during the year, and activities planned for FY-1986

Fusion Safety Program annual report: Fiscal year 1987

International Nuclear Information System (INIS)

Holland, D.F.; Herring, J.S.; Longhurst, G.R.; Lyon, R.E.; Merrill, B.J.; Piet, S.J.

1988-02-01

This report summarizes the Fusion Safety Program major activities in fiscal year 1987. The Idaho National Engineering Laboratory (INEL) is the designated lead laboraotry and EG and G Idaho, Inc., is the prime contractor for this program, which was initiated in 1979. Activities are conducted at the INEL and in participating laboratories including the Hanford Engineering Development Laboratory (HEDL), the Massachusetts Institute of Technology (MIT), and the University of Wisconsin. The technical areas covered in the report include tritium safety, activation product release, reactions involving lithium breeding materials, safety of fusion magnet systems, plasma disruptions, risk assessment methodology, computer codes development for reactor transients, and fusion waste management. Also included in the report is a summary of the safety and environmental analysis and conventional facilities design performed by INEL for the Compact Ignition Tokamak design project, the safety analysis and documentation performed for the Tokamak Ignition/Burn Experimental Reactor design, and the technical support provided to the Environmental Safety and Economics Committee (ESECOM). 42 refs., 17 figs., 4 tabs

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

International Nuclear Information System (INIS)

Oh, Kyemin; Kang, Myoung-suk; Heo, Gyunyoung; Kim, Hyoung-chan

2014-01-01

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

Safety studies on Korean fusion DEMO plant using integrated safety assessment methodology

Energy Technology Data Exchange (ETDEWEB)

Oh, Kyemin; Kang, Myoung-suk [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Heo, Gyunyoung, E-mail: gheo@khu.ac.kr [Kyung Hee University, Youngin-si, Gyeonggi-do 446-701 (Korea, Republic of); Kim, Hyoung-chan [National Fusion Research Institute, Daejeon-si 305-333 (Korea, Republic of)

2014-10-15

Highlights: •The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant. •The concepts of integrated safety assessment methodology (ISAM) that can be applied in addressing regulatory requirements and recognizing safety issues for K-DEMO were emphasized. •Phenomena identification and ranking table (PIRT) was proposed. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional researches. •This work is expected to contribute on the conceptual design of safety features for K-DEMO to design engineers and the guidance for regulatory requirements to licensers. -- Abstract: The purpose of this paper is to suggest methodology that can investigate safety issues and provides a case study for Korean fusion DEMO plant (K-DEMO) as a part of R and D program through the National Fusion Research Institute of Korea. Even though nuclear regulation and licensing framework is well setup due to the operating and design experience of Pressurized Water Reactors (PWRs) since 1970s, the regulatory authority of South Korea has concerns on the challenge of facing new nuclear facilities including K-DEMO due to the differences in systems, materials, and inherent safety feature from conventional PWRs. Even though the follow-up of the ITER license process facilitates to deal with significant safety issues of fusion facilities, a licensee as well as a licenser should identify the gaps between ITER and DEMO in terms of safety issues. First we reviewed the methods of conducting safety analysis for unprecedented nuclear facilities such as Generation IV reactors, particularly very high temperature reactor (VHTR), which is called as integrated safety assessment methodology (ISAM). Second, the analysis for the conceptual design of K-DEMO on the basis of ISAM was conducted. The ISAM consists of five analytical tools to develop the safety requirements from licensee

Inertial fusion reactor designs

International Nuclear Information System (INIS)

Meier, W.

1987-01-01

In this paper, a variety of reactor concepts are proposed. One of the prime concerns is dealing with the x-rays and debris that are emitted by the target. Internal neutron shielding can reduce radiation damage and activation, leading to longer life systems, reduced activation and fewer safety concerns. There is really no consensus on what the best reactor concept is at this point. There has been virtually no chamber technology development to date. This is the flip side of the coin of the separability of the target physics and the reactor design. Since reactor technology has not been required to do target experiments, it's not being developed. Economic analysis of conceptual designs indicates that ICF can be economically competitive with magnetic fusion, fission and fossil plants

Safety vessels for explosive fusion reactor

International Nuclear Information System (INIS)

Mineev, V.

1994-01-01

The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

Fusion safety regulations in the United States: Progress and trends

International Nuclear Information System (INIS)

DeLooper, J.

1994-01-01

This paper explores the issue of regulations as they apply to current and future fusion experimental machines. It addresses fusion regulatory issues, current regulations used for fusion, the Tokamak Fusion Test Reactor experience with regulations, and future regulations to achieve fusion's safety and environmental potential

Safety and environmental aspects of deuterium--tritium fusion power plants: work shop summary

International Nuclear Information System (INIS)

1978-05-01

In September of 1977 a workshop was held on the safety and environmental aspects of fusion power plants to consider potential safety and environmental problems of fusion power plants and to reveal solutions or methods of solving those problems. The objective was to promote incorporation of safety and environmental protection into reactor design, thereby reducing the expense and delay of backfitting safety systems after reactor designs are complete. A dialogue was established between fusion reactor designers and safety and environmental researchers. Four topics, each with several subdivisions, were selected for discussion: radiation exposure, accidents, environmental effects, and plant safety. For each topic, discussion focused on the significance of the problem, and adequacy of current technology to solve the problem, design solutions available and research needed to solve the problem

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

Fission-suppressed hybrid reactor: the fusion breeder

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Coops, M.S.

1982-12-01

Results of a conceptual design study of a 233 U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed

Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

Science.gov (United States)

Fillo, J. A.

This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

Fusion Safety Program Annual Report, Fiscal Year 1996

International Nuclear Information System (INIS)

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

1996-12-01

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

Elemental volatility of HT-9 fusion reactor alloy

International Nuclear Information System (INIS)

Henslee, S.P.; Neilson, R.M. Jr.

1985-01-01

The volatility of elemental constituents from HT-9, a ferritic steel, proposed for fusion reactor structures, was investigated. Tests were conducted in flowing air at temperatures from 800 to 1200 0 C for durations of 1 to 20 h. Elemental volatility was calculated in terms of the weight fraction of the element volatilized from the initial alloy; molybdenum, manganese, and nickel were the primary constituents volatilized. Comparisons with elemental volatilities observed for another candidate fusion reactor materials. Primary Candidate Alloy (PCA), an austenitic stainless steel, indicate significant differences between the volatilities of these steels that may impact fusion reactor safety analysis and alloy selection. Scanning electron microscopy and energy dispersive spectrometry were used to investigate the oxide layers formed on HT-9 and to measure elemental contents within these layers

Fusion safety program annual report fiscal year 1997

International Nuclear Information System (INIS)

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

1998-01-01

This report summarizes the major activities of the Fusion Safety Program in FY 1997. The Idaho National Engineering and Environmental Laboratory (INEEL) is the designated lead laboratory, and Lockheed Martin Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in FY 1979 to perform research and develop data needed to ensure safety in fusion facilities. Activities include experiments, analysis, code development and application, and other forms of research. These activities are conducted at the INEEL, different DOE laboratories, and other institutions. The technical areas covered in this report include chemical reactions and activation product release, tritium safety, 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) project. Work done for ITER this year has focused on developing the needed information for the Non-site Specific Safety Report (NSSR-2)

Fusion safety program annual report fiscal year 1997

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R.; Anderl, R.A.; Cadwallader, L.C. [and others

1998-01-01

This report summarizes the major activities of the Fusion Safety Program in FY 1997. The Idaho National Engineering and Environmental Laboratory (INEEL) is the designated lead laboratory, and Lockheed Martin Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in FY 1979 to perform research and develop data needed to ensure safety in fusion facilities. Activities include experiments, analysis, code development and application, and other forms of research. These activities are conducted at the INEEL, different DOE laboratories, and other institutions. The technical areas covered in this report include chemical reactions and activation product release, tritium safety, 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) project. Work done for ITER this year has focused on developing the needed information for the Non-site Specific Safety Report (NSSR-2).

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2002-01-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

Mirror fusion reactors

International Nuclear Information System (INIS)

Carlson, G.A.; Moir, R.W.

1978-01-01

We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

Mirror fusion reactors

International Nuclear Information System (INIS)

Anon.

1978-01-01

Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

Individual dose due to radioactivity accidental release from fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Ni, Muyi, E-mail: muyi.ni@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Wei, Shiping [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

2017-04-05

Highlights: • Conservative early dose of different unit fusion radioactivity release were assessed. • Data of accident level in INES for fusion reactor were proposed. • Method of environmental restoration time after fusion accident was proposed. • The maximum possible accident level for ITER like fusion reactor is 6. • We need 34–52 years to live after the fusion hypothetical accident. - Abstract: As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1 kg HTO and 1000 kg dust release) and 34–52 years for case 2 (1 kg HTO and 10kg–100 kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation.

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2001-01-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the behaviour of fusion reactor materials and components during and after irradiation. Ongoing projects include: the study of the mechanical behaviour of structural materials under neutron irradiation; the investigation of the characteristics of irradiated first wall material such as beryllium; the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; and the study of dismantling and waste disposal strategy for fusion reactors. Progress and achievements in these areas in 2000 are discussed

Fusion reactors: physics and technology. Annual progress report

International Nuclear Information System (INIS)

Conn, R.W.

1983-08-01

Fusion reactors are designed to operate at full power and generally at steady state. Yet experience shows the load variations, licensing constraints, and frequent sub-system failures often require a plant to operate at fractions of rated power. The aim of this study has been to assess the technology problems and design implications of startup and fractional power operation on fusion reactors. The focus of attention has been tandem mirror reactors (TMR) and we have concentrated on the plasma and blanket engineering for startup and fractional power operation. In this report, we first discuss overall problems of startup, shutdown and staged power operation and their influence on TMR design. We then present a detailed discussion of the plasma physics associated with TMR startup and various means of achieving staged power operation. We then turn to the issue of instrumentation and safety controls for fusion reactors. Finally we discuss the limits on transient power variations during startup and shutdown of Li 17 Pb 83 cooled blankets

Validation of thermal hydraulic codes for fusion reactors safety

International Nuclear Information System (INIS)

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

2006-01-01

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

Materials for fusion reactors

International Nuclear Information System (INIS)

Ehrlich, K.; Kaletta, D.

1978-03-01

The following report describes five papers which were given during the IMF seminar series summer 1977. The purpose of this series was to discuss especially the irradiation behaviour of materials intended for the first wall of future fusion reactors. The first paper deals with the basic understanding of plasma physics relating to the fusion reactor and presents the current state of art of fusion technology. The next two talks discuss the metals intended for the first wall and structural components of a fusion reactor. Since 14 MeV neutrons play an important part in the process of irradiation damage their role is discussed in detail. The question which machines are presently available to simulate irradiation damage under conditions similar to the ones found in a fusion reactor are investigated in the fourth talk which also presents the limitations of the different methods of simulation. In this context also discussed is the importance future intensive neutron sources and materials test reactors will have for this problem area. The closing paper has as a theme the review of the present status of research of metallic and non-metallic materials in view of the quite different requirements for different fusion systems; a closing topic is the world supply on rare materials required for fusion reactors. (orig) [de

EDITORIAL: Safety aspects of fusion power plants

Science.gov (United States)

Kolbasov, B. N.

2007-07-01

This special issue of Nuclear Fusion contains 13 informative papers that were initially presented at the 8th IAEA Technical Meeting on Fusion Power Plant Safety held in Vienna, Austria, 10-13 July 2006. Following recommendation from the International Fusion Research Council, the IAEA organizes Technical Meetings on Fusion Safety with the aim to bring together experts to discuss the ongoing work, share new ideas and outline general guidance and recommendations on different issues related to safety and environmental (S&E) aspects of fusion research and power facilities. Previous meetings in this series were held in Vienna, Austria (1980), Ispra, Italy (1983), Culham, UK (1986), Jackson Hole, USA (1989), Toronto, Canada (1993), Naka, Japan (1996) and Cannes, France (2000). The recognized progress in fusion research and technology over the last quarter of a century has boosted the awareness of the potential of fusion to be a practically inexhaustible and clean source of energy. The decision to construct the International Thermonuclear Experimental Reactor (ITER) represents a landmark in the path to fusion power engineering. Ongoing activities to license ITER in France look for an adequate balance between technological and scientific deliverables and complying with safety requirements. Actually, this is the first instance of licensing a representative fusion machine, and it will very likely shape the way in which a more common basis for establishing safety standards and policies for licensing future fusion power plants will be developed. Now that ITER licensing activities are underway, it is becoming clear that the international fusion community should strengthen its efforts in the area of designing the next generations of fusion power plants—demonstrational and commercial. Therefore, the 8th IAEA Technical Meeting on Fusion Safety focused on the safety aspects of power facilities. Some ITER-related safety issues were reported and discussed owing to their potential

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-02-01

This report describes the engineering conceptual design of Fusion Experimental Reactor (FER) which is to be built as a next generation tokamak machine. This design covers overall reactor systems including MHD equilibrium analysis, mechanical configuration of reactor, divertor, pumped limiter, first wall/breeding blanket/shield, toroidal field magnet, poloidal field magnet, cryostat, electromagnetic analysis, vacuum system, power handling and conversion, NBI, RF heating device, tritium system, neutronics, maintenance, cooling system and layout of facilities. The engineering comparison of a divertor with pumped limiters and safety analysis of reactor systems are also conducted. (author)

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2002-04-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

Radiation environment of fusion experimental reactor

International Nuclear Information System (INIS)

Mori, Seiji; Seki, Yasushi

1988-01-01

Next step device (experimental reactor), which is planned to succeed the large plasma experimental devices such as JT-60, JET and TFTR, generates radiation (neutron + gamma ray) during its operation. Radiation (neutronic) properties of the material are basis for the study on neutron utilization (energy recovery and tritium breeding), material selection (irradiation damage and lifetime evaluation) and radiation safety (personnel exposure and radiation waste). It is necessary, therefore, to predict radiation behaviour in the reactor correctly for the engineering design of the reactor. This report describes the outline of the radiation environment of the reactor based on the information obtained by the neutronic and shielding design calculation of the fusion experimental reactor (FER). (author)

Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Bourham, Mohamed A.; Gilligan, John G.

1999-08-14

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

The fusion reactor

International Nuclear Information System (INIS)

Brennan, M.H.

1974-01-01

Basic principles of the fusion reactor are outlined. Plasma heating and confinement schemes are described. These confinement systems include the linear Z pinch, magnetic mirrors and Tokamaks. A fusion reactor is described and a discussion is given of its environmental impact and its fuel situation. (R.L.)

Safety analysis of fusion reactors pertaining to nuclear incidents and accidents. Final report

International Nuclear Information System (INIS)

Raeder, J.; Weller, A.; Wolf, R.; Jin, X.; Boccaccini, L.V.; Stieglitz, R.; Carloni, D.; Pistner, C.; Herb, J.

2013-11-01

The BfS gave the projekt partners IPP, KIT, Oeko-Institut e. V., and GRS the order to carry out a literature study on the topic of safety of fusion power plants regarding nuclear incidents and accidents. In the framework of this study the actual status of science and technology of the safety concept of fusion power plants should be determined and the applicability of the nuclear safety regulations hitherto developed for nuclear power plants checked. For future commercial fusion power plants today only conceptional designs exist. The most advanced conceptual study for a future fusion power plant is the European Power Plant Conceptual Study (PPCS) from the year 2005, which is based on the tokamak principle. In this study also fundamental aspects of the safety concept of nuclear fusion are treated. Hereby several different conceptual approaches are discussed, which differ among others also in the lay-out approaches relevant for the safety of a facility like for instance the choice of the breeding concept or the materials for the blanket/divertor structure and the coolants. The safety concept of nuclear fusion is oriented on safety concepts for facilities with radioactive inventory. It is based on the concept of tiered safety levels. In order to check whether for the nuclear fusion a safety concept comparable with the nuclear fission at all is necessary, in a first step it was considered, which consequences are possible at a postulated release o large parts of the radioactive inventory of a fusion power plant. Such a worst-case scenario was compared with a corresponding, postulated release of large parts of the radioactive inventory of a nuclear power plant. As scale hereby served the radiological criterion, at the transgression of which in the environment of the facility an evacuation would be necessary. In a next step the transferability of the safety concept of the tiered safety levels of nuclear technology to the fusion was checked. Beside events transferable from

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

International Nuclear Information System (INIS)

1976-01-01

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

Integrity of the first wall in fusion reactors

International Nuclear Information System (INIS)

Kurihara, Ryoichi

2004-07-01

Future fusion power reactors DREAM and A-SSTR2, which have been conceptually designed in the Japan Atomic Energy Research Institute, use the SiC/SiC composite material as the first wall of the blanket because of its characteristics of high heat-resistance and low radiation material. DEMO reactor, which was conceptually designed in 2001, uses the low activation ferritic steel as the first-wall material of the blanket. The problems in the thermal structural design of the plasma facing component such as the blanket first wall and the divertor plate which receives very high heat flux were examined in the design of the fusion power reactors. Compact high fusion power reactor must give high heat flux and high-speed neutron flux from the plasma to the first wall and the divertor plate. In this environmental situation, the micro cracks should be generated in material of the first wall. Structural integrity of the first wall would be very low during the operation of the reactor, if those micro-cracks grow in a crack having significant size by the fatigue or the creep. The crack penetration in the first wall can be a factor which threatens the safety of the fusion power reactor. This paper summarizes the problems on the structural integrity in the first wall made of the SiC/SiC composite material or the ferritic steel. (author)

Safety analysis and evaluation methodology for fusion systems

International Nuclear Information System (INIS)

Fujii-e, Y.; Kozawa, Y.; Namba, C.

1987-03-01

Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)

Fusion reactor development: A review

International Nuclear Information System (INIS)

Anon.

1983-01-01

This paper is a review of the current prospects for fusion reactor development based upon the present status in plasma physics research, fusion technology development and reactor conceptual design for the tokamak magnetic confinement concept. Recent advances in tokamak plasma research and fusion technology development are summarized. The direction and conclusions of tokamak reactor conceptual design are discussed. The status of alternate magnetic confinement concept research is reviewed briefly. A feasible timetable for the development of fusion reactors is presented

Fusion reactor radioactive waste management

International Nuclear Information System (INIS)

Kaser, J.D.; Postma, A.K.; Bradley, D.J.

1976-01-01

Quantities and compositions of non-tritium radioactive waste are estimated for some current conceptual fusion reactor designs, and disposal of large amounts of radioactive waste appears necessary. Although the initial radioactivity of fusion reactor and fission reactor wastes are comparable, the radionuclides in fusion reactor wastes are less hazardous and have shorter half-lives. Areas requiring further research are discussed

Review of recent japanese activities on tritium accountability in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Fukada, Satoshi, E-mail: sfukada@nucl.kyushu-u.ac.jp [Dept. Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-Koen, Kasuga, 816-8580 (Japan); Oya, Yasuhisa [College of Science, Academic Institute, Shizuoka University, 836 Otani, Suruga-ku, Shizuoka 422-8529 (Japan); Hatano, Yuji [Hydrogen Isotope Research Center, Organization for Promotion Research, University of Toyama, 3190 Gofuku, Toyama 930-8555 (Japan)

2016-12-15

Highlights: • Review of Japanese tritium-safety research is given from several viewpoints. • The keywords are tritium accountability and self sufficiency. • Tritium-relating history, tritium facilities and legal regulation are introduced. - Abstract: After introduction of Japanese history or recent topics on tritium (T)-relating research and T-handling capacity in facilities or universities, present activities on T engineering research in Japan are summarized in short in terms of T accountability on safety. The term of safety includes wide processes from T production, assay, storing, confinement, transfer through safety handling finally to shipment of its waste. In order to achieve reliable operation of fusion reactors, several unit processes included in the T cycle of fusion reactors are investigated. Especially, the following recent advances are focused: T retention in plasma facing materials, emergency detritiation system including fire case, T leak through metal tube walls, oxide coating and water detritiation. Strict control, storing and accurate measurement are especially demanded for T accountability depending on various molecular species. Since kg-order T of vaporable radioisotope (RI) will be handled in a fuel cycle or breeding system of a fusion reactor, the accuracy of <0.1% is demanded far over the conventional technology status. Necessity to control T balance within legal restrictions is always kept in mind for operation of the future reactor.

Review of recent japanese activities on tritium accountability in fusion reactors

International Nuclear Information System (INIS)

Fukada, Satoshi; Oya, Yasuhisa; Hatano, Yuji

2016-01-01

Highlights: • Review of Japanese tritium-safety research is given from several viewpoints. • The keywords are tritium accountability and self sufficiency. • Tritium-relating history, tritium facilities and legal regulation are introduced. - Abstract: After introduction of Japanese history or recent topics on tritium (T)-relating research and T-handling capacity in facilities or universities, present activities on T engineering research in Japan are summarized in short in terms of T accountability on safety. The term of safety includes wide processes from T production, assay, storing, confinement, transfer through safety handling finally to shipment of its waste. In order to achieve reliable operation of fusion reactors, several unit processes included in the T cycle of fusion reactors are investigated. Especially, the following recent advances are focused: T retention in plasma facing materials, emergency detritiation system including fire case, T leak through metal tube walls, oxide coating and water detritiation. Strict control, storing and accurate measurement are especially demanded for T accountability depending on various molecular species. Since kg-order T of vaporable radioisotope (RI) will be handled in a fuel cycle or breeding system of a fusion reactor, the accuracy of <0.1% is demanded far over the conventional technology status. Necessity to control T balance within legal restrictions is always kept in mind for operation of the future reactor.

Reactor parameters for European economic, safety and environmental studies

International Nuclear Information System (INIS)

Hancox, R.; Cooke, P.I.H.; Spears, W.R.

1990-01-01

Parameter sets for five 1200 MW e tokamak reactors were developed for the European Study Group on the Environmental, Safety-related and Economic Potential of Fusion Power, showing today's perception of the range of reactors likely to be available as a result of the Commission's fusion programme. On the basis of the cost of generating electricity, relative to a fission reactor, a reference set was chosen and endorsed by the Group for further studies including that on the environmental impact of fusion power. Key physics and technology parameters for the reference reactor are compared with values used in the ITER design, and with those from American studies. (author)

A proposed safety assurance method and its application to the fusion experimental reactor

International Nuclear Information System (INIS)

Okazaki, T.; Seki, Y.; Inabe, T.; Aoki, I.

1995-01-01

Importance categorization and hazard identification methods have been proposed for a fusion experimental reactor. A parameter, the system index, is introduced in the categorization method. The relative importance of systems with safety functions can be classified by the largeness of the system index and whether or not the system acts as a boundary for radioactive materials. This categorization can be used as the basic principle in determining structure design assessment, seismic design criteria etc. For the hazard identification the system time energy matrix is proposed, where the time and spatial distributions of hazard energies are used. This approach is formulated more systematically than an ad-hoc identification of hazard events and it is useful to select design basis events which are employed in the assessment of safety designs. (orig.)

Situation and role of industrial fields in nuclear fusion reactor development

International Nuclear Information System (INIS)

Suzuki, Gen-ichi

1983-01-01

Japan Atomic Industrial Forum (JAIF) established the nuclear fusion technical committee in October, 1980, and has investigated the attitude of industrial fields in progressing nuclear fusion research and development and the measures to cooperate with national development plans. Corresponding to the new long term plan and the establishment of the basic policy for nuclear fusion research and development by Atomic Energy Commission of Japan in June, 1982, JAIF has settled the policy on the situation and role of industrial fields. In this report, first the necessity of firmly grasping the position of nuclear fusion research in atomic energy development is described, next, the present status of the research and development in Japan is reported, and it is mentioned that the role of manufacturers in reinforcing engineering has become more important in industrial fields. In the stage of the construction of a nuclear fusion reactor, the experiences in the engineering safety in fission reactors, environmental safety and system engineering will be utilized. Japanese industrial fields feature that they have made larger cooperation with national projects even in the research and development stage as compared to foreign countries. When the plan of next phase system will be promoted in the future, the cooperating methods in the past should be evaluated, investigated and improved, and the experiences in fast breeder reactors and advanced heavy water reactors should be referred to. Finally, the problems and the countermeasures in nuclear fusion development are described. (Wakatsuki, Y.)

Directions for improved fusion reactors

International Nuclear Information System (INIS)

Krakowski, R.A.; Miller, R.L.; Delene, J.G.

1986-01-01

Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. A generic fusion physics/engineering/costing model is used to provide a quantiative basis for these arguments for specific fusion concepts

Final Report: Safety of Plasma-Facing Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

International Nuclear Information System (INIS)

Bourham, Mohamed A.; Gilligan, John G.

1999-01-01

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m 2 over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER

Fusion reactors - types - problems

International Nuclear Information System (INIS)

Schmitter, K.H.

1979-07-01

A short account is given of the principles of fusion reactions and of the expected advantages of fusion reactors. Descriptions are presented of various Tokamak experimental devices being developed in a number of countries and of some mirror machines. The technical obstacles to be overcome before a fusion reactor could be self-supporting are discussed. (U.K.)

Advanced materials: The key to attractive magnetic fusion power reactors

International Nuclear Information System (INIS)

Bloom, E.E.

1992-01-01

Fusion is one of the most attractive central station power sources from the viewpoint of potential safety and environmental impact characteristics. Studies also indicate that fusion can be economically competitive with other options such as fission reactors and fossil-fired power stations. However, to achieve this triad of characteristics we must develop advanced materials with properties tailored for performance in the various fusion reactor systems. This paper discusses the desired characteristics of materials and the status of materials technology in four critical areas: (1) structural material for the first wail and blanket (FWB), (2) plasma-facing materials, (3) materials for superconducting magnets, and (4) ceramics for electrical and structural applications

Advanced materials - the key to attractive magnetic fusion power reactors

International Nuclear Information System (INIS)

Bloom, E.E.

1992-01-01

Fusion is one of the most attractive central station power sources from the viewpoint of potential safety and environmental impact characteristics. Studies also indicate that fusion can be economically competitive with other options such as fission reactors and fossil-fired power stations. However, to achieve this triad of characteristics we must develop advanced materials with properties tailored for performance in the various fusion reactor systems. This paper discusses the desired characteristics of materials and the status of materials technology in four critical areas: (1) structural materials for the first wall and blanket (FWB), (2) plasmafacing materials, (3) materials for superconducting magnets, and (4) ceramics for electrical and structural applications. (author)

ITER, the 'Broader Approach', a DEMO fusion reactor

International Nuclear Information System (INIS)

Janeschitz, G.; Bahm, W.

2007-01-01

Fusion is a very promising future energy option, which is characterized by almost unlimited fuel reserves, favourable safety features and environmental sustainability. The aim of the worldwide fusion research is a fusion power station which imitates the process taking place in the sun and thus gains energy from the fusion of light atomic nuclei. The experimental reactor ITER which will be built in Cadarache, France, marks a breakthrough in the worldwide fusion research: For the first time an energy multiplication factor of at least 10 will be achieved, the factor by which the fusion power exceeds the external plasma heating. Partners in this project are the European Union, Japan, the Russian Federation, USA, China, South Korea and India as well as Brazil as associated partner. The facility is supposed to demonstrate a long burning, reactor-typical plasma and to test techniques such as plasma heating, plasma confinement by superconducting magnets, fuel cycle as well as energy transition, tritium breeding and remote handling technologies. The next step beyond ITER will be the demonstration power station DEMO which requires further developments in order to create the basis for its design and construction. The roadmap to fusion energy is described. It consists of several elements which are needed to develop the knowledge required for a commercial fusion reactor. The DEMO time schedule depends on the efforts in terms of personnel and budget resources the society is willing to invest in fusion taking into account the long term energy supply and its environmental impact. (orig.)

Compact fusion reactors

CERN Multimedia

CERN. Geneva

2015-01-01

Fusion research is currently to a large extent focused on tokamak (ITER) and inertial confinement (NIF) research. In addition to these large international or national efforts there are private companies performing fusion research using much smaller devices than ITER or NIF. The attempt to achieve fusion energy production through relatively small and compact devices compared to tokamaks decreases the costs and building time of the reactors and this has allowed some private companies to enter the field, like EMC2, General Fusion, Helion Energy, Lawrenceville Plasma Physics and Lockheed Martin. Some of these companies are trying to demonstrate net energy production within the next few years. If they are successful their next step is to attempt to commercialize their technology. In this presentation an overview of compact fusion reactor concepts is given.

Fusion reactor materials

International Nuclear Information System (INIS)

Rowcliffe, A.F.; Burn, G.L.; Knee', S.S.; Dowker, C.L.

1994-02-01

This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Tritium in fusion reactor components

International Nuclear Information System (INIS)

Watson, J.S.; Fisher, P.W.; Talbot, J.B.

1980-01-01

When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

The French-German common safety approach for future reactors

International Nuclear Information System (INIS)

Birkhofer, A.; Chevet, P.F.

1995-01-01

A common safety approach has been defined for future electronuclear plants in the framework of the French-German European Pressurised water Reactor (EPR) project. Improvements in the domain of containment are required in future reactors conception to prevent any risk of core fusion under high and low pressure. Another objective is to reduce significantly the radioactive releases due to other accidents in order to reduce spatial and temporal environmental and human protection procedures. Protection against external aggressions (plane fall, explosions, earthquakes,..), prevention of pipe rupture in the primary circuits, limitation of hydrogen production in the case of water-zirconium complete reaction, cooling of the reactor in the case of core fusion, and radiologic consequences of accidents are the main points discussed by the French-German safety authorities to define the common safety standards of the EPR project. (J.S.)

Possible fusion reactor

International Nuclear Information System (INIS)

Yoshikawa, S.

1976-05-01

A scheme to improve performance characteristics of a tokamak-type fusion reactor is proposed. Basically, the tokamak-type plasma could be moved around so that the plasma could be heated by compression, brought to the region where the blanket surrounds the plasma, and moved so as to keep wall loading below the acceptable limit. This idea should be able to help to economize a fusion reactor

Fusion reactor materials

International Nuclear Information System (INIS)

Anon.

1977-01-01

The following topics are briefly discussed: (1) surface blistering studies on fusion reactor materials, (2) TFTR design support activities, (3) analysis of samples bombarded in-situ in PLT, (4) chemical sputtering effects, (5) modeling of surface behavior, (6) ion migration in glow discharge tube cathodes, (7) alloy development for irradiation performance, (8) dosimetry and damage analysis, and (9) development of tritium migration in fusion devices and reactors

Generic magnetic fusion reactor cost assessment

International Nuclear Information System (INIS)

Sheffield, J.

1985-01-01

The Fusion Energy Division of the Oak Ridge National Laboratory discusses ''generic'' magnetic fusion reactors. The author comments on DT burning magnetic fusion reactor models being possibly operational in the 21st century. Representative parameters from D-T reactor studies are given, as well as a shematic diagram of a generic fusion reactor. Values are given for winding pack current density for existing and future superconducting coils. Topics included are the variation of the cost of electricity (COE), the dependence of the COE on the net electric power of the reactor, and COE formula definitions

The European Fusion Energy Research Programme towards the realization of a fusion demonstration reactor

International Nuclear Information System (INIS)

Gasparotto, M.; Laesser, R.

2006-01-01

Since its inception, the European Fusion Programme has been orientated towards the establishment of the knowledge base needed for the definition of a reactor to be used for power production. Its ultimate goal is then to demonstrate the scientific and the technological feasibility of fusion power while incorporating the assessment of the safety, environmental, social and economic features of this type of energy source. At present, the JET device, the largest tokamak in the world, and the other medium-sized experimental machines are contributing essentially to the basic scientific phase of this development path. Their successful operation greatly contributed to support the design basis of ITER, the next step in fusion, which will aim to demonstrate the scientific and technical feasibility of fusion power production by achieving extended D-T burning plasma operation. Following ITER, the conception and construction of the DEMO device is planned. DEMO will be a demonstration power plant which will be the first fusion device to generate a significant amount of electrical power from fusion. This paper describes the status of fusion research and the European strategy for achievement of the ultimate goal of construction of a prototype reactor. (author)

Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

International Nuclear Information System (INIS)

Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

1997-01-01

Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

Review of fusion DEMO reactor study

International Nuclear Information System (INIS)

Seki, Yasushi

1996-01-01

Fusion DEMO Reactor is defined and the Steady State Tokamak Reactor (SSTR) concept is introduced as a typical example of a DEMO reactor. Recent DEMO reactor studies in Japan and abroad are introduced. The DREAM Reactor concept is introduced as an ultimate target of fusion research. (author)

Environmental aspects of fusion reactors 1985

International Nuclear Information System (INIS)

Casini, G.; Ponti, C.; Rocco, P.

1986-01-01

The aspects of the environmental impact as expected from future fusion reactors are reviewed. The radioactive inventories consist in tritium and neutron-induced radioactivity in the structures. An analysis is performed of the radioactive releases from the different plant's systems in normal and accident conditions and typical emissions to the ambient are defined. Information is given on the waste management problems. Two appendixes give general information on tritium and safety guidelines

Second preliminary design of JAERI experimental fusion reactor (JXFR)

International Nuclear Information System (INIS)

Sako, Kiyoshi; Tone, Tatsuzo; Seki, Yasushi; Iida, Hiromasa; Yamato, Harumi

1979-06-01

Second preliminary design of a tokamak experimental fusion reactor to be built in the near future has been performed. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics radiation shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel recirculating system, reactor cooling and tritium recovery systems and maintenance scheme. Safety analyses of the reactor system have been also performed. This paper gives a brief description of the design as of January, 1979. The feasibility study of raising the power density has been also studied and is shown as appendix. (author)

Advanced nuclear reactor and nuclear fusion power generation

International Nuclear Information System (INIS)

2000-04-01

This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

Radioisotope production in fusion reactors

International Nuclear Information System (INIS)

Engholm, B.A.; Cheng, E.T.; Schultz, K.R.

1986-01-01

Radioisotope production in fusion reactors is being investigated as part of the Fusion Applications and Market Evaluation (FAME) study. /sup 60/Co is the most promising such product identified to date, since the /sup 60/Co demand for medical and food sterilization is strong and the potential output from a fusion reactor is high. Some of the other radioisotopes considered are /sup 99/Tc, /sup 131/l, several Eu isotopes, and /sup 210/Po. Among the stable isotopes of interest are /sup 197/Au, /sup 103/Rh and Os. In all cases, heat or electricity can be co-produced from the fusion reactor, with overall attractive economics

Present status of design, research and development of nuclear fusion reactors and problems

International Nuclear Information System (INIS)

1983-04-01

Seven years have elapsed since the publication of ''Progress of nuclear fusion research and perspective toward the development of power reactors'' by the Atomic Energy Society of Japan in August, 1976. During this period, the research and development of nuclear fusion have changed from plasma physics to reactor technology, being conscious of the realization of fusion reactors. There are the R project in the Institute of Plasma Physics, Nagoya University, and the design and construction of JT-60 in Japan Atomic Energy Research Institute, to put it concretely. Now the research and development taking the economical efficiency into account are adopted. However, the type of fusion reactors is not reduced to tokamak type, accordingly the research and development to meet the diverse possibilities are forwarded. The progress of tokamak reactor research, core plasma design, nuclear design and shielding design, thermal structure design, the design of superconducting magnets, disassembling and repair, safety, economical efficiency, the conceptual design of other types than tokamak and others are reported. (Kako, I.)

What have fusion reactor studies done for you today?

International Nuclear Information System (INIS)

Kulchinski, G.L.

1985-01-01

The University of Wisconsin examines the fusion program and puts into perspective what return is being made on investments in fusion reactor studies. Illustations show financial support for fusion research from the four major programs, FY'82 expenditures on fusion research, and the total expenditures on fusion research since 1951. Topics discussed include the estimated number of scientists conducting fusion research, the conceptual design study of a fusion reactor, scoping study of a reactor, the chronology of fusion reactor design studies, published fusion reactor studies 1967-1983, conceptual fusion reactor design studies, STARFIRE reference design, MARS central cell, HYLIFE reaction chamber, and selected contributions of reactor design studies to base programs

Fusion reactors and the environment

International Nuclear Information System (INIS)

Hancox, R.

1990-04-01

Fusion power, based on the nuclear fusion of light elements to yield a net gain of energy, has the potential to extend the world's resources in a way which is environmentally attractive. Nevertheless, the easiest route to fusion - the reaction between deuterium and tritium - involves hazards from the use of tritium and the neutron activation of the structural materials. These hazards have been considered on the basis of simple conceptual reactor designs, both in relation to normal operation and decommissioning and to potential accident situations. Results from several studies are reviewed and suggest that fusion reactors appear to have an inherently lower environmental impact than fission reactors. However, the realization of this potential has yet to be demonstrated. (author)

Coatings for fusion reactor environments

International Nuclear Information System (INIS)

Mattox, D.M.

1979-01-01

The internal surfaces of a tokamak fusion reactor control the impurity injection and gas recycling into the fusion plasma. Coating of internal surfaces may provide a desirable and possibly necessary design flexibility for achieving the temperatures, ion densities and containment times necessary for net energy production from fusion reactions to take place. In this paper the reactor environments seen by various componentare reviewed along with possible materials responses. Characteristics of coating-substrate systems, important to fusion applications, are delineated and the present status of coating development for fusion applications is reviewed. Coating development for fusion applications is just beginning and poses a unique and important challenge for materials development

Pulsed fusion reactors

International Nuclear Information System (INIS)

1975-01-01

This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

Fusion reactor cost reductions by employing non-nuclear grade components

International Nuclear Information System (INIS)

Bourque, R.F.; Maya, I.; Schultz, K.R.; Sonn, D.L.; Wise, R.K.

1987-09-01

The Cascade inertial confinement fusion reactor fits the requirements of low radioactive inventories and inherent safety and is therefore a candidate for non-nuclear construction throughout. This reactor consists of a rotating blanket of ceramic granules that absorb the energy from D-T target explosions occurring along the rotational axis. Laser energy is beamed in axially from both ends. Two cost estimates were made for an 815 MWe Cascade power plant. One was based on an ''all conventional'' plant, which is constructed and costed using well-established, conventional fossil power plant methods. The second was a ''nuclear plus conventional'' design, constructed and costed using a combination of fossil and fission reactor plant methods and standards that would be typical of advanced fission reactors. The total capital requirements for the ''all conventional'' construction plant were estimated in 1985 dollars at $1490 M, including indirect costs. Similarly, the ''nuclear plus conventional'' construction plant was estimated at $1940 M. The savings of $450 M (23%) represents strictly the difference between Cascade ICF power plants designed and constructed to nuclear safety-related requirements versus all non-nuclear. This example clearly shows that, if fusion plants can take advantage of low activation materials and inherent safety features to eliminate the need for nuclear-related expenses, then such plants may have economic advantages over nuclear-grade systems. 13 refs., 1 fig., 5 tabs

Fusion reactor wastes

International Nuclear Information System (INIS)

Young, J.R.

1976-01-01

The fusion reactor currently is being developed as a clean source of electricity with an essentially infinite source of fuel. These reactors are visualized as using a fusion reaction to generate large quantities of high temperature energy which can be used as process heat or for the generation of electricity. The energy would be created primarily as the kinetic energy of neutrons or other reaction products. Neutron energy could be converted to high-temperature heat by moderation and capture of the neutrons. The energy of other reaction products could be converted to high-temperature heat by capture, or directly to electricity by direct conversion electrostatic equipment. An analysis to determine the wastes released as a result of operation of fusion power plants is presented

Technical issues in fusion reactors

International Nuclear Information System (INIS)

Rohatgi, V.K.; Vijayan, T.

1989-01-01

In this paper the issues in fusion reactor technology are examined. Rapid progress in fusion technology research in recent years can be attributed to the advances in various technologies. The commercial generation of fusion power greatly depends on the evolution and improvements in these technologies. With better understanding of plasma physics, fusion reactor designs are becoming more and more realistic and comprehensive. It is now possible to compare various concepts within the framework of established technologies. The technological issues needing better understanding and solutions to problem areas are identified. Various instabilities and energy losses are major problem areas. Extensive developments in reactor-relevant advanced materials, compact and powerful superconducting magnets, high-power systems, and plasma heating drivers need to be undertaken and emphasized

Polymer materials for fusion reactors

International Nuclear Information System (INIS)

Yamaoka, H.

1993-01-01

The radiation-resistant polymer materials have recently drawn much attention from the viewpoint of components for fusion reactors. These are mainly applied to electrical insulators, thermal insulators and structural supports of superconducting magnets in fusion reactors. The polymer materials used for these purposes are required to withstand the synergetic effects of high mechanical loads, cryogenic temperatures and intense nuclear radiation. The objective of this review is to summarize the anticipated performance of candidate materials including polymer composites for fusion magnets. The cryogenic properties and the radiation effects of polymer materials are separately reviewed, because there is only limited investigation on the above-mentioned synergetic effects. Additional information on advanced polymer materials for fusion reactors is also introduced with emphasis on recent developments. (orig.)

Intelligible seminar on fusion reactors. (12) Next step toward the realization of fusion reactors. Future vision of fusion energy research and development

International Nuclear Information System (INIS)

Okano, Kunihiko; Kurihara, Kenichi; Tobita, Kenji

2006-01-01

In the last session of this seminar the progress of research and development for the realization of fusion reactors and future vision of fusion energy research and development are summarized. The some problems to be solved when the commercial fusion reactors would be realized, (1) production of deuterium as the fuel, (2) why need the thermonuclear reactors, (3) environmental problems, and (4) ITER project, are described. (H. Mase)

Survey of fusion reactor technology

International Nuclear Information System (INIS)

Chung, M.K.; Kang, H.D.; Oh, Y.K.; Lee, K.W.; In, S.Y.; Kim, Y.C.

1983-01-01

The present object of the fusion research is to accomplish the scientific break even by the year of 1986. In view of current progress in the field of Fusion reactor development, we decided to carry out the conceptual design of Tokamak-type fusion reactor during the year of 82-86 in order to acquire the principles of the fusion devices, find the engineering problems and establish the basic capabilities to develop the key techniques with originality. In this year the methods for calculating the locations of the poloidal coils and distribution of the magnetic field, which is one of the most essential and complicated task in the fusion reactor design works, were established. Study on the optimization of the design method of toroidal field coil was also done. Through this work, we established the logic for the design of the toroidal field coil in tokamak and utilize this technique to the design of small compact tokamak. Apart from the development work as to the design technology of tokamak, accelerating column and high voltage power supply (200 KVDC, 100 mA) for intense D-T neutron generator were constructed and now beam transport systems are under construction. This device will be used to develop the materials and the components for the tokamak fusion reactor. (Author)

Some safety considerations in laser-controlled thermonuclear reactors. Final report

International Nuclear Information System (INIS)

Botts, T.E.; Breton, D.; Chan, C.K.; Levy, S.I.; Sehnert, M.; Ullman, A.Z.

1978-07-01

A major objective of this study was to identify potential safety questions for laser controlled thermonuclear reactors. From the safety viewpoint, it does not appear that the actual laser controlled thermonuclear reactor conceptual designs present hazards very different than those of magnetically confined fusion reactors. Some aspects seem beneficial, such as small lithium inventories, and the absence of cryogenic devices, while other aspects are new, for example the explosion of pressure vessels and laser hazards themselves. Major aspects considered in this report include: (a) general safety considerations, (b) tritium inventories, (c) system behavior during loss of flow accidents, and (d) safety considerations of laser related penetrations

Review of the safety concept for fusion reactor concepts and transferability of the nuclear fission regulation to potential fusion power plants

Energy Technology Data Exchange (ETDEWEB)

Raeder, Juergen; Weller, Arthur; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik (IPP), Garching (Germany); Jin, Xue Zhou; Boccaccini, Lorenzo V.; Stieglitz, Robert; Carloni, Dario [Karlsruher Institute fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany); Pistner, Christoph [Oeko-Institut e.V., Darmstadt (Germany); Herb, Joachim [Gesellschaft fuer Anlagen- und Reaktorsicherheit, Koeln (Germany)

2016-01-15

This paper summarizes the current state of the art in science and technology of the safety concept for future fusion power plants (FPPs) and examines the transferability of the current nuclear fission regulation to the concepts of future fusion power plants. At the moment there exist only conceptual designs of future fusion power plants. The most detailed concepts with regards to safety aspects were found in the European Power Plant Conceptual Study (PPCS). The plant concepts discussed in the PPCS are based on magnetic confinement of the plasma. The safety concept of fusion power plants, which has been developed during the last decades, is based on the safety concepts of installations with radioactive inventories, especially nuclear fission power plants. It applies the concept of defence in depth. However, there are specific differences between the implementations of the safety concepts due to the physical and technological characteristics of fusion and fission. It is analysed whether for fusion a safety concept is required comparable to the one of fission. For this the consequences of a purely hypothetical release of large amounts of the radioactive inventory of a fusion power plant and a fission power plant are compared. In such an event the evacuation criterion outside the plant is exceeded by several orders of magnitude for a fission power plant. For a fusion power plant the expected radiological consequences are of the order of the evacuation criterion. Therefore, a safety concept is also necessary for fusion to guarantee the confinement of the radioactive inventory. The comparison between the safety concepts for fusion and fission shows that the fundamental safety function ''confinement of the radioactive materials'' can be transferred directly in a methodical way. For a fusion power plant this fundamental safety function is based on both, physical barriers as well as on active retention functions. After the termination of the fusion

Hydrogen production in fusion reactors

Science.gov (United States)

Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

1993-11-01

As one of the methods of innovative energy production in fusion reactors (that do not include a conventional turbine-type generator), the efficient use of fusion-reactor radiation and semiconductors to supply clean fuel in the form of hydrogen gas is studied. Taking the reactor candidates such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a plant system concept are investigated.

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)

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Pulse Star inertial confinement fusion reactor

International Nuclear Information System (INIS)

Blink, J.A.; Hogan, W.J.

1985-01-01

Pulse Star is a pool-type ICF reactor that emphasizes low cost and high safety levels. The reactor consists of a vacuum chamber (belljar) submerged in a compact liquid metal (Li 17 Pb 83 or lithium) pool which also contains the heat exchangers and liquid metal pumps. The shielding efficiency of the liquid metal pool is high enough to allow hands-on maintenance of (removed) pumps and heat exchangers. Liquid metal is allowed to spray through the 5.5 m radius belljar at a controlled rate, but is prohibited from the target region by a 4 m radius mesh first wall. The wetted first wall absorbs the fusion x-rays and debris while the spray region absorbs the fusion neutrons. The mesh allows vaporized liquid metal to blow through to the spray region where it can quickly cool and condense. Preliminary calculations show that a 2 m thick first wall could handle the mechanical (support, buckling, and x-ray-induced hoop) loads. Wetting and gas flow issues are in an initial investigation stage

Core fusion accidents in nuclear power reactors. Knowledge review

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

Fusion reactor materials

International Nuclear Information System (INIS)

Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

1980-01-01

Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2000-01-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2000-07-01

SCK-CEN's research and development programme on fusion reactor materials includes: (1) the study of the mechanical behaviour of structural materials under neutron irradiation (including steels, inconel, molybdenum, chromium); (2) the determination and modelling of the characteristics of irradiated first wall materials such as beryllium; (3) the detection of abrupt electrical degradation of insulating ceramics under high temperature and neutron irradiation; (4) the study of the dismantling and waste disposal strategy for fusion reactors.; (5) a feasibility study for the testing of blanket modules under neutron radiation. Main achievements in these topical areas in the year 1999 are summarised.

Hydrogen production in fusion reactors

International Nuclear Information System (INIS)

Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

1993-11-01

As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated. (author)

Hydrogen production in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

1993-11-01

As one of methods of innovative energy production in fusion reactors without having a conventional turbine-type generator, an efficient use of radiation produced in a fusion reactor with utilizing semiconductor and supplying clean fuel in a form of hydrogen gas are studied. Taking the candidates of reactors such as a toroidal system and an open system for application of the new concepts, the expected efficiency and a concept of plant system are investigated. (author).

Fusion Safety Program annual report: Fiscal year 1986

International Nuclear Information System (INIS)

Holland, D.F.; Merrill, B.J.; Herring, J.S.; Piet, S.J.; Longhurst, G.R.

1987-06-01

This report summarizes the Fusion Safety Program's (FSP) major activities in fiscal year 1986. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory, and EG and G Idaho, Inc., is the prime contractor for FSP, which was initiated in 1979. Activities are conducted at the INEL and in participating facilities, including the Hanford Engineering Development Laboratory (HEDL), the Massachusetts Institute of Technology (MIT), and the University of Wisconsin. The technical areas covered in this report include tritium safety, activation product release, reactions involving lithium breeding materials, safety of fusion magnet systems, plasma disruption, risk assessment methodology, and computer code development for reactor transients. Contributions to the Technical Planning Activity (TPA) and the ''white paper'' study by the Environmental, Safety,and Economics Committee (ESECOM) are summarized. The report also includes a summary of the safety and environmental analysis and documentation performed by the INEL for the Compact Ignition Tokamak (CIT) design project

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

Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor

International Nuclear Information System (INIS)

Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

2015-01-01

As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. - Highlights: • Dynamic tritium concentration in the air and human body evaluated on the time scale. • Different intake forms and relevant radiation dose assessed to find out the potential relationships. • HT and HTO release amount limits for arbitrary accidents proposed for the fusion reactor according to dose limit

Fusion systems engineering

International Nuclear Information System (INIS)

Anon.

1977-01-01

Summaries of research are included for each of the following topics: (1) fusion reactor systems studies, (2) development of blanket processing technology for fusion reactors, (3) safety studies of fusion concepts, (4) the MACK/MACKLIB system for nuclear response functions, and (5) energy storage and power supply systems for fusion reactors

Vacuum engineering for fusion research and fusion reactors

International Nuclear Information System (INIS)

Pittenger, L.C.

1976-01-01

The following topics are described: (1) surface pumping by cryogenic condensation, (2) operation of large condensing cryopumps, (3) pumping for large fusion experiments, and (4) vacuum technology for fusion reactors

Fusion systems engineering

International Nuclear Information System (INIS)

Anon.

1978-01-01

Research during this report period has covered the following areas: (1) fusion reactor systems studies, (2) development of blanket processing technology for fusion reactors, (3) safety studies of fusion concepts, (4) MACKLIB-IV, a new library of nuclear response functions, (5) energy storage and power supply requirements for commercial fusion reactors, (6) blanket/shield design evaluation for commercial fusion reactors, and (7) cross section measurements, evaluations, and techniques

ITER: a technology test bed for a fusion reactor

International Nuclear Information System (INIS)

Huguet, M.; Green, B.J.

1996-01-01

The ITER Project aims to establish nuclear fusion as an energy source that has potential safety and environmental advantages, and to develop the technologies required for a fusion reactor. ITER is a collaborative project between the European Union, Japan, the Russian Federation and the United States of America. During the current phase of the Project, an R and D programme of about 850 million dollars is underway to develop the technologies required for ITER. This technological effort should culminate in the construction of the components and systems of the ITER machine and its auxiliaries. The main areas of technological development include the first wall and divertor technology, the blanket technology and tritium breeding, superconducting magnet technology, pulsed power technology and remote handling. ITER is a test bed and an essential step to establish the technology of future fusion reactors. Many of the ITER technologies are of potential interest to other fields and their development is expected to benefit the industries involved. (author)

Experimental and analytical investigations to air and steam ingress into the vacuum vessel of fusion reactors

International Nuclear Information System (INIS)

Kruessenberg, A.K.

1996-12-01

The basic fusion safety objective is the development of fusion power plants with features that protect individuals, society and the environment by establishing and maintaining an effective defence against radiological and other hazards. The most important specific principle is the establishment of three sequential levels of defence, characterized in priority order by prevention, protection and mitigation. The safety conscious selection of materials as one prevention feature gives the basis for the work described in this report. In order to protect the metallic first wall of fusion reactors from direct interaction with the plasma an extra armour is foreseen. Carbon offers the features low atomic number, high melting point, high thermal conductivity and good mechanical stability up to high temperatures making it to a favourite armour material. Looking on the safety behaviour of fusion reactors it has to be noted that carbon is unstable against oxidizing media like oxygen and steam at high temperatures und carbon has a high sorption capacity for radiologically important tritium. And tritium used as intermediate fuel in the actual reactor concepts is the one form radioactivity is present in fusion reactors. Accidents like loss of vacuum (LOVA) will lead to an air ingress into the vacuum vessel, oxidation of the hot carbon and a partial mobilization of the sorbed tritium. In a similar manner loss of coolant into vacuum (LOCIV) will lead to a water/steam ingress into the vacuum vessel, also accompanied by carbon oxidation and tritium release. (orig.)

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

Status of fusion reactor concept development in Japan

International Nuclear Information System (INIS)

Tsuji-Iio, Shunji

1996-01-01

Fusion power reactor studies in Japan based on magnetic confinement schemes are reviewed. As D-T fusion reactors, a steady-state tokamak reactor (SSTR) was proposed and extensively studied at the Japan Atomic Energy Research Institute (JAERI) and an inductively operated day-long tokamak reactor (IDLT) was proposed by a group at the University of Tokyo. The concept of a drastically easy maintenance (DREAM) tokamak reactor is being developed at JAERI. A high-field tokamak reactor with force-balanced coils as a volumetric neutron source is being studied by our group at Tokyo Institute of Technology. The conceptual design of a force-free helical reactor (FFHR) is under way at the National Institute for Fusion Science. A design study of a D- 3 He field-reversed configuration (FRC) fusion reactor called ARTEMIS was conducted by the FRC fusion working group of research committee of lunar base an lunar resources. (author)

Fusion reactor problems

International Nuclear Information System (INIS)

Carruthers, R.

It is pointed out that plasma parameters for a fusion reactor have been fairly accurately defined for many years, and the real plasma physics objective must be to find the means of achieving and maintaining these specifiable parameters. There is good understanding of the generic technological problems: breading blankets and shields, radiation damage, heat transfer and methods of magnet design. The required plasma parameters for fusion self-heated reactors are established at ntausub(E) approximately 2.10 14 cm -3 sec, plasma radius 1.5 to 3 m, wall loading 5 to 10 MW cm -2 , temperature 15 keV. Within this model plasma control by quasi-steady burn as a key problem is studied. It is emphasized that the future programme must interact more closely with engineering studies and should concentrate upon research which is relevant to reactor plasmas. (V.P.)

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

New materials in nuclear fusion reactors

International Nuclear Information System (INIS)

Iwata, Shuichi

1988-01-01

In the autumn of 1987, the critical condition was attained in the JET in Europe and Japanese JT-60, thus the first subject in the physical verification of nuclear fusion reactors was resolved, and the challenge to the next attainment of self ignition condition started. As the development process of nuclear fusion reactors, there are the steps of engineering, economical and social verifications after this physical verification, and in respective steps, there are the critical problems related to materials, therefore the development of new materials must be advanced. The condition of using nuclear fusion reactors is characterized by high fluence, high thermal flux and strong magnetic field, and under such extreme condition, the microscopic structures of materials change, and they behave much differently from usual case. The subjects of material development for nuclear fusion reactors, the material data base being built up, the materials for facing plasma and high thermal flux, first walls, blanket structures, electric insulators and others are described. The serious effect of irradiation and the rate of defect inducement must be taken in consideration in the structural materials for nuclear fusion reactors. (Kako, I.)

Dynamic evaluation of environmental impact due to tritium accidental release from the fusion reactor.

Science.gov (United States)

Nie, Baojie; Ni, Muyi; Jiang, Jieqiong; Wu, Yican

2015-10-01

As one of the key safety issues of fusion reactors, tritium environmental impact of fusion accidents has attracted great attention. In this work, the dynamic tritium concentrations in the air and human body were evaluated on the time scale based on accidental release scenarios under the extreme environmental conditions. The radiation dose through various exposure pathways was assessed to find out the potential relationships among them. Based on this work, the limits of HT and HTO release amount for arbitrary accidents were proposed for the fusion reactor according to dose limit of ITER. The dynamic results aim to give practical guidance for establishment of fusion emergency standard and design of fusion tritium system. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

Small mirror fusion reactors

International Nuclear Information System (INIS)

Carlson, G.A.; Schultz, K.R.; Smith, A.C. Jr.

1978-01-01

Basic requirements for the pilot plants are that they produce a net product and that they have a potential for commercial upgrade. We have investigated a small standard mirror fusion-fission hybrid, a two-component tandem mirror hybrid, and two versions of a field-reversed mirror fusion reactor--one a steady state, single cell reactor with a neutral beam-sustained plasma, the other a moving ring field-reversed mirror where the plasma passes through a reaction chamber with no energy addition

Present status of inertial confinement fusion reactor design

International Nuclear Information System (INIS)

Mima, Kunioki; Ido, Shunji; Nakai, Sadao.

1986-01-01

Since inertial nuclear fusion reactors do not require high vacuum and high magnetic field, the structure of the reactor cavity becomes markedly simple as compared with tokamak type fusion reactors. In particular, since high vacuum is not necessary, liquid metals such as lithium and lead can be used for the first wall, and the damage of reactor structures by neutrons can be prevented. As for the core, the energy efficiency of lasers is not very high, accordingly it must be designed so that the pellet gain due to nuclear fusion becomes sufficiently high, and typically, the gain coefficient from 100 to 200 is necessary. In this paper, the perspective of pellet gain, the plan from the present status to the practical reactors, and the conceptual design of the practical reactors are discussed. The plan of fuel ignition, energy break-even and high gain by the implosion mode, of which the uncertain factor due to uneven irradiation and instability was limited to the minimum, was clarified. The scenario of the development of laser nuclear fusion reactors is presented, and the concept of the reactor system is shown. The various types of nuclear fusion-fission hybrid reactors are explained. As for the design of inertial fusion power reactors, the engineering characteristics of the core, the conceptual design, water fall type reactors and DD fuel reactors are discussed. (Kako, I.)

Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

1995-01-01

Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

Analysis of Confinement Strategies for a Tokamak Fusion Reactor

Energy Technology Data Exchange (ETDEWEB)

Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

2001-01-15

The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established.

Analysis of Confinement Strategies for a Tokamak Fusion Reactor

International Nuclear Information System (INIS)

Girard, Christian; Gaillard, Jean-Philippe; Marbach, Gabriel; Cambi, Gilio; Cook, Ian; Johansson, Lise-Lotte; Meyder, Rainer; Mustoe, Julian; Pinna, Tonio

2001-01-01

The Safety and Environmental Assessment of Fusion Power (SEAFP) was performed in the framework of the European fusion program, whose results have already been published. The European Commission decided to continue this program for some identified issues that required development. One of these issues was the analysis and specification of the containment concepts that minimize accidental releases to the environment.To perform such an assessment, a methodology was followed to identify the most challenging accidental sequences in terms of containment integrity.The results of the accident selection and analysis that were performed during the extension of the SEAFP-2 program are given. Preliminary recommendations for the definition of a confinement strategy for tokamak fusion reactors are established

Advanced fusion reactor

International Nuclear Information System (INIS)

Tomita, Yukihiro

2003-01-01

The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p- 6 Li and p- 11 B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D- 3 He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D- 3 He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of 3 He per a year. On the other hand, 1 million tons of 3 He is estimated to be in the moon. The 3 He of about 10 23 kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

Survey of the laser-solenoid fusion reactor

International Nuclear Information System (INIS)

Amherd, N.A.

1975-09-01

This report surveys the prospects for a laser-solenoid fusion reactor. A sample reactor and scaling laws are used to describe the concept's characteristics. Experimental results are reviewed, and the laser and magnet technologies that undergird the laser-solenoid concept are analyzed. Finally, a systems analysis of fusion power reactors is given, including a discussion of direct conversion and fusion-fission effects, to ascertain the system attributes of the laser-solenoid configuration

Inherent/passive safety for fusion

International Nuclear Information System (INIS)

Piet, S.J.

1986-06-01

The concept of inherent or passive passive safety for fusion energy is explored, defined, and partially quantified. Four levels of safety assurance are defined, which range from true inherent safety to passive safety to protection via active engineered safeguard systems. Fusion has the clear potential for achieving inherent or passive safety, which should be an objective of fusion research and design. Proper material choice might lead to both inherent safety and high mass power density, improving both safety and economics. When inherent safety is accomplished, fusion will be well on the way to achieving its ultimate potential and to be truly different and superior

Radiolytic production of chemical fuels in fusion reactor systems

Energy Technology Data Exchange (ETDEWEB)

Fish, J D

1977-06-01

Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered.

Radiolytic production of chemical fuels in fusion reactor systems

International Nuclear Information System (INIS)

Fish, J.D.

1977-06-01

Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered

Advances in laser solenoid fusion reactor design

International Nuclear Information System (INIS)

Steinhauer, L.C.; Quimby, D.C.

1978-01-01

The laser solenoid is an alternate fusion concept based on a laser-heated magnetically-confined plasma column. The reactor concept has evolved in several systems studies over the last five years. We describe recent advances in the plasma physics and technology of laser-plasma coupling. The technology advances include progress on first walls, inner magnet design, confinement module design, and reactor maintenance. We also describe a new generation of laser solenoid fusion and fusion-fission reactor designs

Introduction to magnetic fusion reactor design

International Nuclear Information System (INIS)

Watanabe, Kenji

1988-01-01

Trend of the tokamak reactor design works so far carried out is reviewed, and method of conceptual design for commercial fusion reactor is critically considered concerning the black-box conpepts. System-framework of the engineering of magnetic fusion (commercial) reactor design is proposed as four steps. Based on it the next design studies are recommended in parallel approaches for making real-overcome of reactor material problem, from the view point of technological realization and not from the economical one. Real trials are involved. (author)

Tritium resources available for fusion reactors

Science.gov (United States)

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

2018-02-01

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

The need and prospects for improved fusion reactors

International Nuclear Information System (INIS)

Krakowski, R.A.; Hagenson, R.L.; Miller, R.L.

1986-01-01

Conceptual fusion reactor studies over the past 10-15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100-200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed

Activation product transport in fusion reactors

International Nuclear Information System (INIS)

Klein, A.C.; Vogelsang, W.F.

1984-01-01

Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. A computer code, RAPTOR, has been developed to determine the transport of these products in fusion reactor coolant/tritium breeding materials. Without special treatment, it is likely that fusion reactor power plant operators could experience dose rates as high as 8 rem per hour around a number of plant components after only a few years of operation. (orig.)

Advanced fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Tomita, Yukihiro [National Inst. for Fusion Science, Toki, Gifu (Japan)

2003-04-01

The main subjects on fusion research are now on D-T fueled fusion, mainly due to its high fusion reaction rate. However, many issues are still remained on the wall loading by the 14 MeV neutrons. In the case of D-D fueled fusion, the neutron wall loading is still remained, though the technology related to tritium breeding is not needed. The p-{sup 6}Li and p-{sup 11}B fueled fusions are not estimated to be the next generation candidate until the innovated plasma confinement technologies come in useful to achieve the high performance plasma parameters. The fusion reactor of D-{sup 3}He fuels has merits on the smaller neutron wall loading and tritium handling. However, there are difficulties on achieving the high temperature plasma more than 100 keV. Furthermore the high beta plasma is needed to decrease synchrotron radiation loss. In addition, the efficiency of the direct energy conversion from protons coming out from fusion reaction is one of the key parameters in keeping overall power balance. Therefore, open magnetic filed lines should surround the plasma column. In this paper, we outlined the design of the commercial base reactor (ARTEMIS) of 1 GW electric output power configured by D-{sup 3}He fueled FRC (Field Reversed Configuration). The ARTEMIS needs 64 kg of {sup 3}He per a year. On the other hand, 1 million tons of {sup 3}He is estimated to be in the moon. The {sup 3}He of about 10{sup 23} kg are to exist in gaseous planets such as Jupiter and Saturn. (Y. Tanaka)

Maintenance of fission and fusion reactors. 10. workshop on fusion reactor engineering

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

This report contains copies of OHP presented at the title meeting. The presented topics are as follows, maintenance of nuclear power plants and ITER, exchange of shroud in BWR type reactors, deterioration of fission and fusion reactor materials, standards of pressure vessels, malfunction diagnosis method with neural network. (J.P.N.)

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

Investigation of materials for fusion power reactors

Science.gov (United States)

Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

2014-06-01

The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

Thermal-hydraulic limitations on water-cooled fusion reactor components

International Nuclear Information System (INIS)

Cha, Y.S.; Misra, B.

1986-01-01

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

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

Some safety considerations of liquid lithium as a fusion breeder material

International Nuclear Information System (INIS)

Jeppson, D.W.; Muhlestein, L.D.

1986-01-01

Test results and conclusions are presented for the reaction of steam with a high temperature lithium pool and for the reaction of high temperature lithium spray with a nitrogen atmosphere. The reactions are characterized and evaluated in regard to the potential for mobilization of radioactive species associated with the liquid breeder under postulated fusion reactor accident conditions. These evaluations include measured lithium temperature responses, atmosphere temperature and pressure responses, gas consumption and generation, aerosol quantities and particle size characterization, and potentially radioactive species releases. Conclusions are made as to the consequences of these safety considerations for the use of lithium as a fusion reactor breeder material

Advances in fusion reactor design

International Nuclear Information System (INIS)

Baker, C.C.

1987-01-01

The author addresses the tokamak as a power reactor. Contrary to popular opinion, there are still a few people that think a tokamak might make a good fusion power reactor. In thinking about advances in fusion reactor design, in the U.S., at least, that generally means advances relevant to the Starfire design. He reviews some of the features of Starfire. Starfire is the last major study done of the tokamak as a reactor in this country. It is now over eight years old in the sense that eight years ago was really the time in which major decisions were made as to its features. Starfire was a tokamak with a major radius of seven meters, about twice the linear dimensions of a machine like TIBER

The tritium and the controlled fusion reactors

International Nuclear Information System (INIS)

Leger, D.; Rouyer, J.L.

1986-04-01

It is shown how tritium is used how it is circulating in a fusion reactor. The great functions of tritium circuits are detailed: reprocessing of burnt gases, reprocessing of gases coming from neutral injectors, reprocessing from gaseous wastes, detritiation of cooling fluids. Current technologic developments are quoted. Then tritium confinement and containment, in normal or accidental situations, are displayed. Limitation devices of effluents and release for normal operating (noticeably the reprocessing systems of atmosphere) and safety and protection systems in case of accident are described [fr

Reactor concepts for laser fusion

International Nuclear Information System (INIS)

Meier, W.R.; Maniscalco, J.A.

1977-07-01

Scoping studies were initiated to identify attractive reactor concepts for producing electric power with laser fusion. Several exploratory reactor concepts were developed and are being subjected to our criteria for comparing long-range sources of electrical energy: abundance, social costs, technical feasibility, and economic competitiveness. The exploratory concepts include: a liquid-lithium-cooled stainless steel manifold, a gas-cooled graphite manifold, and fluidized wall concepts, such as a liquid lithium ''waterfall'', and a ceramic-lithium pellet ''waterfall''. Two of the major reactor vessel problems affecting the technical feasibility of a laser fusion power plant are: the effects of high-energy neutrons and cyclical stresses on the blanket structure and the effects of x-rays and debris from the fusion microexplosion on the first-wall. The liquid lithium ''waterfall'' concept is presented here in more detail as an approach which effectively deals with these damaging effects

Composites as structural materials in fusion reactors

International Nuclear Information System (INIS)

Megusar, J.

1989-01-01

In fusion reactors, materials are used under extreme conditions of temperature, stress, irradiation, and chemical environment. The absence of adequate materials will seriously impede the development of fusion reactors and might ultimately be one of the major difficulties. Some of the current materials problems can be solved by proper design features. For others, the solution will have to rely on materials development. A parallel and balanced effort between the research in plasma physics and fusion-related technology and in materials research is, therefore, the best strategy to ultimately achieve economic, safe, and environmentally acceptable fusion. The essential steps in developing composites for structural components of fusion reactors include optimization of mechanical properties followed by testing under fusion-reactor-relevant conditions. In optimizing the mechanical behavior of composite materials, a wealth of experience can be drawn from the research on ceramic matrix and metal matrix composite materials sponsored by the Department of Defense. The particular aspects of this research relevant to fusion materials development are methodology of the composite materials design and studies of new processing routes to develop composite materials with specific properties. Most notable examples are the synthesis of fibers, coatings, and ceramic materials in their final shapes form polymeric precursors and the infiltration of fibrous preforms by molten metals

Space Propulsion via Spherical Torus Fusion Reactor

International Nuclear Information System (INIS)

Williams, Craig H.; Juhasz, Albert J.; Borowski, Stanley K.; Dudzinski, Leonard A.

2003-01-01

A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 204 days, with an initial mass in low Earth orbit of 1630 mt. Engineering conceptual design, analysis, and assessment were performed on all major systems including nuclear fusion reactor, magnetic nozzle, power conversion, fast wave plasma heating, fuel pellet injector, startup/re-start fission reactor and battery, and other systems. Detailed fusion reactor design included analysis of plasma characteristics, power balance and utilization, first wall, toroidal field coils, heat transfer, and neutron/X-ray radiation

Cooling System Design Options for a Fusion Reactor

Science.gov (United States)

Natalizio, Antonio; Collén, Jan; Vieider, Gottfried

1997-06-01

The objective of a fusion power reactor is to produce electricity safely and reliably. Accordingly, the design, objective of the heat transport system is to optimize power production, safety, and reliability. Such an optimization process, however, is constrained by many factors, including, among others: public safety, worker safety, steam cycle efficiency, reliability, and cost. As these factors impose conflicting requirements, there is a need to find an optimum design solution, i.e., one that satisfies all requirements, but not necessarily each requirement optimally. The SEAFP reactor study developed helium-cooled and water-cooled models for assessment purposes. Among other things, the current study demonstrates that neither model offers an optimum solution. Helium cooling offers a high steam cycle efficiency but poor reliability for the cooling of high heat flux components (divertor and first wall). Alternatively, water cooling offers a low steam cycle efficiency, but reasonable reliability for the cooling of such components. It is concluded that an optimum solution includes helium cooling of low heat flux components and water cooling of high heat flux components. Relative to the SEAFP helium model, this hybrid system enhances safety and reliability, while retaining the high steam cycle efficiency of that model.

Public acceptance of fusion energy and scientific feasibility of a fusion reactor. DREAM (DRastically EAsy Maintenance) tokamak

International Nuclear Information System (INIS)

Nishio, Satoshi

1998-01-01

If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)

Fusion reactor start-up without an external tritium source

Energy Technology Data Exchange (ETDEWEB)

Zheng, S., E-mail: Shanliang.Zheng@ccfe.ac.uk; King, D.B.; Garzotti, L.; Surrey, E.; Todd, T.N.

2016-02-15

Highlights: • Investigated the feasibility (including plasma physics, neutronics and economics) of starting a fusion reactor from running pure D–D fusion reactor to gradually move towards the D–T operation. • Proposed building up tritium from making use of neutrons generated by D–D fusion reactions. • Studied plasma physics feasibility for pure D–D operation and provided consistent fusion power and neutron yield in the plasma with different mixture of deuterium and tritium. • Discussed the economics aspect for operating a pure D–D fusion reactor towards a full-power D–T fusion reactor. - Abstract: It has long been recognised that the shortage of external tritium sources for fusion reactors using D–T, the most promising fusion fuel, requires all such fusion power plants (FPP) to breed their own tritium. It is also recognised that the initial start-up of a fusion reactor will require several kilograms of tritium within a scenario in which radioactive decay, ITER and subsequent demonstrator reactors are expected to have consumed most of the known tritium stockpile. To circumvent this tritium fuel shortage and ultimately achieve steady-state operation for a FPP, it is essential to first accumulate sufficient tritium to compensate for loss due to decay and significant retention in the materials in order to start a new FPP. In this work, we propose to accumulate tritium starting from D–D fusion reactions, since D exists naturally in water, and to gradually build up the D–T plasma targeted in fusion reactor designs. There are two likely D–D fusion reaction channels, (1) D + D → T + p, and (2) D + D → He3 + n. The tritium can be generated via the reaction channel ‘(1)’ and the 2.45 MeV neutrons from ‘(2)’ react with lithium-6 in the breeding blanket to produce more tritium to be fed back into plasma fuel. Quantitative evaluations are conducted for two blanket concepts to assess the feasibility and suitability of this approach to FPP

Review of mirror fusion reactor designs

International Nuclear Information System (INIS)

Bender, D.J.

1977-01-01

Three magnetic confinement concepts, based on the mirror principle, are described. These mirror concepts are summarized as follows: (1) fusion-fission hybrid reactor, (2) tandem mirror reactor, and (3) reversed field mirror reactor

Tritium problems in fusion reactor systems

International Nuclear Information System (INIS)

Hickman, R.G.

1975-01-01

A brief introduction is given to the role tritium will play in the development of fusion power. The biological and worldwide environmental behavior of tritium is reviewed. The tritium problems expected in fusion power reactors are outlined. A few thoughts on tritium permeation and recent results for tritium cleanup and CT 4 accumulation are presented. Problems involving the recovery of tritium from the breeding blanket in fusion power reactors are also considered, including the possible effect of impurities in lithium blankets and the use of lithium as a regenerable getter pump. (auth)

AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 3

Energy Technology Data Exchange (ETDEWEB)

Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Programm Nukleare Sicherheitsforschung (NUKLEAR); Mull, Thomas [AREVA GmbH, Erlangen (Germany). Nuclear Fusion, HTR and Transverse Issues (PTDH-G); Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

2015-01-15

Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Reactor Operation, Safety: Radiation Protection (Angelika Bohnstedt); - Competence, Innovation, Regulation: Fusion Technology - Optimisation Steps in the ITER Design (Thomas Mull); - Competence, Innovation, Regulation: Education, Expert Knowledge, Knowledge Transfer (Joerg Starflinger). The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 and 12 (2015) and will be covered in further issues of atw.

Fusion systems engineering

International Nuclear Information System (INIS)

Anon.

1977-01-01

Information is given on each of the following topics: (1) fusion reactor systems studies, (2) development of blanket processing technology for fusion reactors, (3) safety studies of CTR concepts, and (4) cross section measurements and techniques

Inertial fusion reactors and magnetic fields

International Nuclear Information System (INIS)

Cornwell, J.B.; Pendergrass, J.H.

1985-01-01

The application of magnetic fields of simple configurations and modest strengths to direct target debris ions out of cavities can alleviate recognized shortcomings of several classes of inertial confinement fusion (ICF) reactors. Complex fringes of the strong magnetic fields of heavy-ion fusion (HIF) focusing magnets may intrude into reactor cavities and significantly affect the trajectories of target debris ions. The results of an assessment of potential benefits from the use of magnetic fields in ICF reactors and of potential problems with focusing-magnet fields in HIF reactors conducted to set priorities for continuing studies are reported. Computational tools are described and some preliminary results are presented

Neutron streaming evaluation for the DREAM fusion power reactor

International Nuclear Information System (INIS)

Seki, Yasushi; Nishio, Satoshi; Ueda, Shuzo; Kurihara, Ryoichi

2000-01-01

Aiming at high degree of safety and benign environmental effect, we have proposed a tokamak fusion reactor concept called DREAM, which stands for DRastically EAsy Maintenance Reactor. The blanket structure of the reactor is made from very low activation SiC/SiC composites and cooled by non-reactive helium gas. High net thermal efficiency of about 50% is realized by 900 C helium gas and high plant availability is possible with simple maintenance scheme. In the DREAM Reactor, neutron streaming is a big problem because cooling pipes with diameter larger than 80 cm are used for blanket heat removal. Neutron streaming through the cooling pipes could cause hot spots in the superconducting magnets adjacent to the cooling pipes to shorten the magnet lifetime or increase cryogenic cooling requirement. Neutron streaming could also activate components such as gas turbine further away from the fusion plasma. The effect of neutron streaming through the helium cooling pipes was evaluated for the two types of cooling pipe extraction scheme. The result of a preliminary calculation indicates the gas turbine activation prohibits personnel access in the case of inboard pipe extraction while with additional shielding measures, limited contact maintenance is possible in the case of outboard extraction. (author)

Advanced nuclear fuel production by using fission-fusion hybrid reactor

International Nuclear Information System (INIS)

Al-Kusayer, T.A.; Sahin, S.; Abdulraoof, M.

1993-01-01

Efforts are made at the College of Engineering, King Saud University, Riyadh to lay out the main structure of a prototype experimental fusion and fusion-fission (hybrid) reactor blanket in cylindrical geometry. The geometry is consistent with most of the current fusion and hybrid reactor design concepts in respect of the neutronic considerations. Characteristics of the fusion chamber, fusion neutrons and the blanket are provided. The studies have further shown that 1 GWe fission-fusion reactor can produce up to 957 kg/year which is enough to fuel five light water reactors of comparable power. Fuel production can be increased further. 29 refs

Summary of the report of the Senior Committee on Environmental, Safety, and Economic Aspects of Magnetic Fusion Energy

International Nuclear Information System (INIS)

Holdren, J.P.; Berwald, D.H.; Budnitz, R.J.

1987-01-01

The Senior Committee on Environmental, Safety, and Economic Aspects of Magnetic Fusion Energy (ESECOM) has assessed magnetic fusion energy's prospects for providing energy with economic, environmental, and safety characteristics that would be attractive compared with other energy sources (mainly fission) available in the year 2015 and beyond. ESECOM gives particular attention to the interaction of environmental, safety, and economic characteristics of a variety of magnetic fusion reactors, and compares them with a variety of fission cases. Eight fusion cases, two fusion-fission hybrid cases, and four fission cases are examined, using consistent economic and safety models. These models permit exploration of the environmental, safety, and economic potential of fusion concepts using a wide range of possible materials choices, power densities, power conversion schemes, and fuel cycles. The ESECOM analysis indicates that magnetic fusion energy systems have the potential to achieve costs-of-electricity comparable to those of present and future fission systems, coupled with significant safety and environmental advantages. 75 refs., 2 figs., 24 tabs

Conceptual design study of quasi-steady state fusion experimental reactor (FER-Q), part 2

International Nuclear Information System (INIS)

1985-12-01

Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. Starting from 1984 FER design is being reviewed and redesigned. This report is a part of the interim report which describes the results obtained in the review and redesign activities in FY 1984. The results of the following design items are included: heating/current drive system, plasma position control, power supply, diagnostics, neutronics, blanket test module, repair and maintenance and safety. (author)

Chemical Hazards and Safety Issues in Fusion Safety Design

International Nuclear Information System (INIS)

Cadwallader, L.C.

2003-01-01

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

Parameter study toward economical magnetic fusion power reactors

International Nuclear Information System (INIS)

Yoshida, Tomoaki; Okano, Kunihiko; Nanahara, Toshiya; Hatayama, Akiyoshi; Yamaji, Kenji; Takuma, Tadashi.

1996-01-01

Although the R and D of nuclear fusion reactors has made a steady progress as seen in ITER project, it has become of little doubt that fusion power reactors require hugeness and enormous amount of construction cost as well as surmounting the physics and engineering difficulties. Therefore, it is one of the essential issues to investigate the prospect of realizing fusion power reactors. In this report we investigated the effects of physics and engineering improvements on the economics of ITER-like steady state tokamak fusion reactors using our tokamak system and costing analysis code. With the results of this study, we considered what is the most significant factor for realizing economical competitive fusion reactors. The results show that with the conventional TF coil maximum field (12T), physics progress in β-value (or Troyon coefficient) has the most considerable effect on the reduction of fusion plant COE (Cost of Electricity) while the achievement of H factor = 2-3 and neutron wall load =∼5MW/m 2 is necessary. The results also show that with the improvement of TF coil maximum field, reactors with a high aspect ratio are economically advantageous because of low plasma current driving power while the improvement of current density in the conductors and yield strength of support structures is indispensable. (author)

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

International Nuclear Information System (INIS)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

Energy Technology Data Exchange (ETDEWEB)

Koch, M.; Kazimi, M.S.

1991-04-01

Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

Trends and developments in magnetic confinement fusion reactor concepts

International Nuclear Information System (INIS)

Baker, C.C.; Carlson, G.A.; Krakowski, R.A.

1981-01-01

An overview is presented of recent design trends and developments in reactor concepts for magnetic confinement fusion. The paper emphasizes the engineering and technology considerations of commercial fusion reactor concepts. Emphasis is placed on reactors that operate on the deuterium/tritium/lithium fuel cycle. Recent developments in tokamak, mirror, and Elmo Bumpy Torus reactor concepts are described, as well as a survey of recent developments on a wide variety of alternate magnetic fusion reactor concepts. The paper emphasizes recent developments of these concepts within the last two to three years

Studies of conceptual spheromak fusion reactors

International Nuclear Information System (INIS)

Katsurai, M.; Yamada, M.

1982-01-01

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A >or approx. 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type operation, (2) the ignited DT reactor, and (3) the ignited catalysed-type DD reactor. With a total wall loading of approx. 4 MW.m -2 , it is found that edge magnetic fields of only approx. 4 T (DT) and approx. 9 T (Cat. DD) are required for ignited reactors of 1 m plasma (minor) radius with output powers in the gigawatt range. An assessment of various schemes of generation, compression and translation of spheromak plasmas is presented. (author)

Survey on the fusion/fission-hybrid-reactors, a literature review

International Nuclear Information System (INIS)

A survey, based on existing literature, of the work being pursued worldwide on fusion - fission (hybrid) reactor systems is presented. Six areas are reviewed: Plasma physics parameters; Blankets concepts; Fuel cycles; Reactor conceptual designs; Safety and environmental problems; System studies and economic perspectives. Attention has been restricted to systems using magnetically confined plasmas, mainly to mirror and Tokamak - type concepts. The aim is to provide sufficient information, even if not exhaustive, on hybrid reactor concepts in order to help understand what may be expected from their possible development and the ways in which hybrids could affect the future energy scenario. Some concluding remarks are made which represent the personal view of the authors only

Radiation shielding for fusion reactors

International Nuclear Information System (INIS)

Santoro, R.T.

2000-01-01

Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

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

Fusion Safety Program annual report, fiscal year 1984

International Nuclear Information System (INIS)

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

1985-06-01

This report summarizes the Fusion Safety Program major activities in fiscal year 1984. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and EG and G Idaho, Inc., is the prime contractor for this program, which was initiated in 1979. A report section titled ''Activities at the INEL'' includes progress reports on the tritium implantation experiment, tritium blanket permeation, volatilization of reactor alloys, plasma disruptions, a comparative blanket safety assessment, transient code development, and a discussion of the INEL's participation in the Tokamak Fusion Core Experiment (TFCX) design study. The report section titled ''Outside Contracts'' includes progress reports on tritium conversion by the Oak Ridge National Laboratory (ORNL), lithium-lead reactions by the Hanford Engineering Development Laboratory (HEDL) and the University of Wisconsin, magnet safety by the Francis Bitter Magnet Laboratory of the Massachusetts Institute of Technology (MIT) and Argonne National Laboratory (ANL), risk assessment by MIT, tritium retention by the University of Virginia, and activation product release by GA Technologies. A list of publications produced during the year and brief descriptions of activities planned for FY-1985 are also included

Present status of fusion reactor materials, 4

International Nuclear Information System (INIS)

Nagasaki, Ryukichi; Shiraishi, Kensuke; Watanabe, Hitoshi; Murakami, Yoshio; Takamura, Saburo

1982-01-01

Recently, the design of fusion reactors such as Intor has been carried out, and various properties that fusion reactor materials should have been clarified. In the Japan Atomic Energy Research Institute, the research and development of materials aiming at a tokamak type experimental fusion reactor are in progress. In this paper, the problems, the present status of research and development and the future plan about the surface materials and structural materials for the first wall, blanket materials and magnet materials are explained. The construction of the critical plasma testing facility JT-60 developed by JAERI has progressed smoothly, and the operation is expected in 1985. The research changes from that of plasma physics to that of reactor technology. In tokamak type fusion reactors, high temperature D-T plasma is contained with strong magnetic field in vacuum vessels, and the neutrons produced by nuclear reaction, charged particles diffusing from plasma and neutral particles by charge exchange strike the first wall. The PCA by improving 316 stainless steel is used as the structural material, and TiC coating techniques are developed. As the blanket material, Li 2 O is studied, and superconducting magnets are developed. (Koko, I.)

Public acceptance of fusion energy and scientific feasibility of a fusion reactor. Design of inductively driven long pulse tokamak reactors: IDLT

International Nuclear Information System (INIS)

Ogawa, Yuichi

1998-01-01

Based on scientific data based adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R and D program, the scientific feasibility of inductively-driven tokamak fusion reactors is studied. A low wall-loading DEMO fusion reactor is designed, which utilizes an austenitic stainless steel in conjunction with significant data bases and operating experiences, since we have given high priority to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the DEMO reactor with the relatively large volume (i.e., major radius of 10 m) is employed, plasma ignition is achievable with a low fusion power of 0.8 GW, and an operation period of 4 - 5 hours is available only with inductive current drive. Disadvantages of pulsed operation in commercial fusion reactors include fatigue in structural materials and the necessity of an energy storage system to compensate the electric power during the dwell time. To overcome these disadvantages, a pulse length is prolonged up to about 10 hours, resulting in the remarkable reduction of the total cycle number to 10 4 during the life of the fusion plant. (author)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Updated comparison of economics of fusion reactors with advanced fission reactors

International Nuclear Information System (INIS)

Delene, J.G.

1990-01-01

The projected cost of electricity (COE) for fusion is compared with that from current and advanced nuclear fission and coal-fired plants. Fusion cost models were adjusted for consistency with advanced fission plants and the calculational methodology and cost factors follow guidelines recommended for cost comparisons of advanced fission reactors. The results show COEs of about 59--74 mills/kWh for the fusion designs considered. In comparison, COEs for future fission reactors are estimated to be in the 43--54 mills/kWh range with coal-fired plant COEs of about 53--69 mills/kWh ($2--3/GJ coal). The principal cost driver for the fusion plants relative to fission plants is the fusion island cost. Although the estimated COEs for fusion are greater than those for fission or coal, the costs are not so high as to preclude fusion's competitiveness as a safe and environmentally sound alternative

Standard mirror fusion reactor design study

International Nuclear Information System (INIS)

Moir, R.W.

1978-01-01

This report covers the work of the Magnetic Fusion Energy Division's reactor study group during FY 1976 on the standard mirror reactor. The ''standard'' mirror reactor is characterized as a steady state, neutral beam sustained, D-T fusioning plasma confined by a Yin-Yang magnetic mirror field. The physics parameters are obtained from the same physics model that explains the 2XIIB experiment. The model assumes that the drift cyclotron loss cone mode occurs on the boundary of the plasma, and that it is stabilized by warm plasma with negligible energy investment. The result of the study was a workable mirror fusion power plant, steady-state blanket removal made relatively simple by open-ended geometry, and no impurity problem due to the positive plasma potential. The Q (fusion power/injected beam power) turns out to be only 1.1 because of loss out the ends from Coulomb collisions, i.e., classical losses. This low Q resulted in 77% of the gross electrical power being used to power the injectors, thereby causing the net power cost to be high. The low Q stimulated an intensive search for Q-enhancement concepts, resulting in the LLL reactor design effort turning to the field reversal mirror and the tandem mirror, each having Q of order 5

Nuclear data needs for fusion reactors

International Nuclear Information System (INIS)

Gohar, Y.

1986-01-01

The nuclear design of fusion components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems. 84 refs., 11 figs., 9 tabs

Neutron dosimetry for radiation damage in fission and fusion reactors

International Nuclear Information System (INIS)

Smith, D.L.

1979-01-01

The properties of materials subjected to the intense neutron radiation fields characteristic of fission power reactors or proposed fusion energy devices is a field of extensive current research. These investigations seek important information relevant to the safety and economics of nuclear energy. In high-level radiation environments, neutron metrology is accomplished predominantly with passive techniques which require detailed knowledge about many nuclear reactions. The quality of neutron dosimetry has increased noticeably during the past decade owing to the availability of new data and evaluations for both integral and differential cross sections, better quantitative understanding of radioactive decay processes, improvements in radiation detection technology, and the development of reliable spectrum unfolding procedures. However, there are problems caused by the persistence of serious integral-differential discrepancies for several important reactions. There is a need to further develop the data base for exothermic and low-threshold reactions needed in thermal and fast-fission dosimetry, and for high-threshold reactions needed in fusion-energy dosimetry. The unsatisfied data requirements for fission reactor dosimetry appear to be relatively modest and well defined, while the needs for fusion are extensive and less well defined because of the immature state of fusion technology. These various data requirements are examined with the goal of providing suggestions for continued dosimetry-related nuclear data research

Advanced fuels for nuclear fusion reactors

International Nuclear Information System (INIS)

McNally, J.R. Jr.

1974-01-01

Should magnetic confinement of hot plasma prove satisfactory at high β (16 πnkT//sub B 2 / greater than 0.1), thermonuclear fusion fuels other than D.T may be contemplated for future fusion reactors. The prospect of the advanced fusion fuels D.D and 6 Li.D for fusion reactors is quite promising provided the system is large, well reflected and possesses a high β. The first generation reactions produce the very active, energy-rich fuels t and 3 He which exhibit a high burnup probability in very hot plasmas. Steady state burning of D.D can ensue in a 60 kG field, 5 m reactor for β approximately 0.2 and reflectivity R/sub mu/ = 0.9 provided the confinement time is about 38 sec. The feasibility of steady state burning of 6 Li.D has not yet been demonstrated but many important features of such systems still need to be incorporated in the reactivity code. In particular, there is a need for new and improved nuclear cross section data for over 80 reaction possibilities

Fusion-fission hybrid as an alternative to the fast breeder reactor

International Nuclear Information System (INIS)

Barrett, R.J.; Hardie, R.W.

1980-09-01

This report compares the fusion-fission hybrid on the plutonium cycle with the classical fast breeder reactor (FBR) cycle as a long-term nuclear energy source. For the purpose of comparison, the current light-water reactor once-through (LWR-OT) cycle was also analyzed. The methods and models used in this study were developed for use in a comparative analysis of conventional nuclear fuel cycles. Assessment areas considered in this study include economics, energy balance, proliferation resistance, technological status, public safety, and commercial viability. In every case the characteristics of all fuel cycle facilities were accounted for, rather than just those of the reactor

Breeder control fusion reactor. Topical interview

Energy Technology Data Exchange (ETDEWEB)

Schlueter, A [Max-Planck-Institut fuer Plasmaphysik, Garching/Muenchen (Germany, F.R.)

1977-09-01

The energy sources of the future are extremely controversial. The consumption of fossil fuel shall decrease during the next decades, because exhaustion of the resources, pollution, increase of CO/sub 2/ in the atmosphere and other reasons. But at present the question it not yet settled which alternative energy system should replace the fossil fuel. First of all nuclear energy in the form of fission reactions seems to come into operation to a larger extent. The next step may be the controlled thermonuclear fusion reaction. Furthermore, a comparison between fusion and fission is given which shows that fusion would bring about less risks than the breeders. An advantage of the fusion reactor would be the fact that the fuel cycle is closed. Unfortunately, the physical questions are not as yet satisfactorily clarified so that one cannot be sure whether a fusion reactor can really be built.

Reactor safety

International Nuclear Information System (INIS)

Butz, H.P.; Heuser, F.W.; May, H.

1985-01-01

The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

Nuclear characteristics of D-D fusion reactor blankets

International Nuclear Information System (INIS)

Nakashima, Hideki; Ohta, Masao

1978-01-01

Fusion reactors operating on deuterium (D-D) cycle are considered to be of long range interest for their freedom from tritium breeding in the blanket. The present paper discusses the various possibilities of D-D fusion reactor blanket designs mainly from the standpoint of the nuclear characteristics. Neutronic and photonic calculations are based on presently available data to provide a basis of the optimal blanket design in D-D fusion reactors. It is found that it appears desirable to design a blanket with blanket/shield (BS) concept in D-D fusion reactors. The BS concept is designed to obtain reasonable shielding characteristics for superconducting magnet (SCM) by using shielding materials in the compact blanket. This concept will open the possibility of compact radiation shield design based on assured technology, and offer the advantage from the system economics point of view. (auth.)

An analysis of the estimated capital cost of a fusion reactor

International Nuclear Information System (INIS)

Hollis, A.A.

1981-06-01

The cost of building a fusion reactor similar to the Culham Conceptual Tokamak reactor Mark IIB is assessed and compared with other published capital costs of fusion and fission reactors. It is concluded that capital-investment and structure-renewal costs for a typical fusion reactor as presently conceived are likely to be higher than for thermal-fission reactors. (author)

An analysis of the estimated capital cost of a fusion reactor

International Nuclear Information System (INIS)

Hollis, A.A.; Evans, L.S.

1981-01-01

The cost of building a fusion reactor similar to the Culham Conceptual Tokamak reactor Mark IIB is assessed and compared with other published capital costs of fusion and fission reactors. It is concluded that capital-investment and structure-renewal costs for a typical fusion reactor as presently conceived are likely to be higher than for thermal-fission reactors. (author)

Confinement inertial fusion. Power reactors of nuclear fusion by lasers

International Nuclear Information System (INIS)

Velarde, G.; Ahnert, C.; Aragones, J.M.; Leira, G; Martinez-Val, J.M.

1980-01-01

The energy crisis and the need of the nuclear fusion energy are analized. The nuclear processes in the laser interation with the ablator material are studied, as well as the thermohydrodinamic processes in the implossion, and the neutronics of the fusion. The fusion reactor components are described and the economic and social impact of its introduction in the future energetic strategies.(author)

Synergies in the design and development of fusion and generation IV fission reactors

International Nuclear Information System (INIS)

Bogusch, E.; Carre, F.; Knebel, J.; Aoto, K.

2007-01-01

Future fusion reactors or systems and Generation IV fission reactors are designed and developed in worldwide programmes mostly involving the same partners to investigate and assess their potential for realisation and contribution to meet the future energy needs beyond 2030. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core face similar design issues and development needs. Therefore the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactors or systems will be designed for helium and liquid metal cooling and higher temperatures similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches might create synergistic design and development programmes. Therefore an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in support of common technologies. (orig.)

Synergies in the design and development of fusion and generation IV fission reactors

International Nuclear Information System (INIS)

Bogusch, E.; Carre, F.; Knebel, J.U.; Aoto, K.

2008-01-01

Future fusion reactor and Generation IV fission reactor systems are designed and developed in worldwide programmes to investigate and assess their potential for realisation and contribution to the future energy needs beyond 2030 mostly involving the same partners. Huge scientific and financial effort is necessary to meet these objectives. First programmes have been launched in Generation IV International Forum (GIF) for fission and in the Broader Approach for fusion reactor system development. Except for the physics basis for the energy source, future fusion and fission reactors, in particular those with fast neutron core, face similar design issues and development needs. Therefore, the call for the identification of synergies became evident. Beyond ITER cooled by water, future fusion reactor systems will be designed for high-temperature helium and liquid metal cooling but also water including supercritical water and molten salt similar to those proposed for some of the six fission reactor concepts in GIF with their diverse coolants. Beside materials developments which are not discussed in this paper, design and performance of components and systems related to the diverse coolants including lifetime and maintenance aspects might offer significant potentials for synergies. Furthermore, the use of process heat for applications in addition to electricity production as well as their safety approaches can create synergistic design and development programmes. Therefore, an early identification of possible synergies in the relevant programmes should be endorsed to minimise the effort for future power plants in terms of investments and resources. In addition to a general overview of a possible synergistic work programme which promotes the interaction between fusion and fission programmes towards an integrated organisation of their design and R and D programmes, some specific remarks will be given for joint design tools, numerical code systems and joint experiments in

SOLASE: a conceptual laser fusion reactor design

International Nuclear Information System (INIS)

Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

1977-12-01

The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion reactors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

Prospects for toroidal fusion reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.D.

1994-01-01

Work on the International Thermonuclear Experimental Reactor (ITER) tokamak has refined understanding of the realities of a deuterium-tritium (D-T) burning magnetic fusion reactor. An ITER-like tokamak reactor using ITER costs and performance would lead to a cost of electricity (COE) of about 130 mills/kWh. Advanced tokamak physics to be tested in the Toroidal Physics Experiment (TPX), coupled with moderate components in engineering, technology, and unit costs, should lead to a COE comparable with best existing fission systems around 60 mills/kWh. However, a larger unit size, ∼2000 MW(e), is favored for the fusion system. Alternative toroidal configurations to the conventional tokamak, such as the stellarator, reversed-field pinch, and field-reversed configuration, offer some potential advantage, but are less well developed, and have their own challenges

Fusion-Fission hybrid reactors and nonproliferation

International Nuclear Information System (INIS)

Greenspan, E.

1984-09-01

New options for the development of the nuclear energy economy which might become available by a successful development of fusion-breeders or fusion-fission hybrid power reactors, identified and their nonproliferative attributes are discussed. The more promising proliferation-resistance ettributes identified include: (1) Justification for a significant delay in the initiation of fuel processing, (2) Denaturing the plutonium with 238 Pu before its use in power reactors of any kind, and (3) Making practical the development of denatured uranium fuel cycles and, in particular, denaturing the uranium with 232 U. Fuel resource utilization, time-table and economic considerations associated with the use of fusion-breeders are also discussed. It is concluded that hybrid reactors may enable developing a nuclear energy economy which is more proliferation resistant than possible otherwise, whileat the same time, assuring high utilization of t he uranium and thorium resources in an economically acceptable way. (author)

Status of fusion reactor blanket design

International Nuclear Information System (INIS)

Smith, D.L.; Sze, D.K.

1986-02-01

The recent Blanket Comparison and Selection Study (BCSS), which was a comprehensive evaluation of fusion reactor blanket design and the status of blanket technology, serves as an excellent basis for further development of blanket technology. This study provided an evaluation of over 130 blanket concepts for the reference case of electric power producing, DT fueled reactors in both Tokamak and Tandem Mirror (TMR) configurations. Based on a specific set of reactor operating parameters, the current understanding of materials and blanket technology, and a uniform evaluation methodology developed as part of the study, a limited number of concepts were identified that offer the greatest potential for making fusion an attractive energy source

Fusion Reactor Safety Research program. Annual report, FY-80

International Nuclear Information System (INIS)

Crocker, J.G.; Cohen, S.

1981-06-01

The report is in three sections. Outside contracts includes a report of newly-started study at the General Atomic Company to consider safety implications of low-activation materials, portions of two papers from ongoing work at PNL and ANL, reports of the lithium spill work at HEDL, the LITFIRE code development at MIT, and risk assessment at MIT, all of which are an expansion of FY-79 outside contracts. EG and G Activities includes adaptations of four papers of ongoing work in transient code development, tritium system risk assessment, heat transfer and fluid flow analysis, and fusion safety data base. Program Plan Development includes the Executive Summary of the Plan, which was completed in FY-80, and is accompanied by a list of publications and a brief outline of proposed FY-81 activities to be based on the Program Plan

Fusion reactor design studies: standard accounts for cost estimates

International Nuclear Information System (INIS)

Schulte, S.C.; Willke, T.L.; Young, J.R.

1978-05-01

The fusion reactor design studies--standard accounts for cost estimates provides a common format from which to assess the economic character of magnetically confined fusion reactor design concepts. The format will aid designers in the preparation of design concept costs estimates and also provide policymakers with a tool to assist in appraising which design concept may be economically promising. The format sets forth a categorization and accounting procedure to be used when estimating fusion reactor busbar energy cost that can be easily and consistently applied. Reasons for developing the procedure, explanations of the procedure, justifications for assumptions made in the procedure, and the applicability of the procedure are described in this document. Adherence to the format when evaluating prospective fusion reactor design concepts will result in the identification of the more promising design concepts thus enabling the fusion power alternatives with better economic potential to be quickly and efficiently developed

Prospects for spheromak fusion reactors

International Nuclear Information System (INIS)

Fowler, T.K.; Hua, D.D.

1995-01-01

The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on physics principles confirmed in CTX experiments in many respects. Most uncertain was the energy confinement time and the role of magnetic turbulence inherent in the concept. In this paper, a one-dimensional model of heat confinement, calibrated by CTX, predicts negligible heat loss by magnetic turbulence at reactor scale

Neutronics issues in fusion-fission hybrid reactor design

International Nuclear Information System (INIS)

Liu Chengan

1995-01-01

The coupled neutron and γ-ray transport equations and nuclear number density equations, and its computer program systems concerned in fusion-fission hybrid reactor design are briefly described. The current status and focal point for coming work of nuclear data used in fusion reactor design are explained

Self-sustaining nuclear pumped laser-fusion reactor experiment

International Nuclear Information System (INIS)

Boody, F.P.; Choi, C.K.; Miley, G.H.

1977-01-01

The features of a neutron feedback nuclear pumped (NFNP) laser-fusion reactor equipment were studied with the intention of establishing the feasibility of the concept. The NFNP laser-fusion concept is compared schematically to electrically pumped laser fusion. The study showed that, once a method of energy storage has been demonstrated, a self-sustaining fusion-fission hybrid reactor with a ''blanket multiplication'' of two would be feasible using nuclear pumped Xe F* excimer lasers having efficiencies of 1 to 2 percent and D-D-T pellets with gains of 50 to 100

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

Symbiosis of near breeder HTR's with hybrid fusion reactors

International Nuclear Information System (INIS)

Seifritz, W.

1978-07-01

In this contribution to INFCE a symbiotic fusion/fission reactor system, consisting of a hybrid beam-driven micro-explosion fusion reactor (HMER) and associated high-temperature gas-cooled reactors (HTR) with a coupled fuel cycle, is proposed. This system is similar to the well known Fast Breeder/Near Breeder HTR symbiosis except that the fast fission breeder - running on the U/Pu-cycle in the core and the axial blankets and breeding the surplus fissile material as U-233 in its radial thorium metal or thorium oxide blankets - is replaced by a hybrid micro-explosion DT fusion reactor

Heat transfer in inertial confinement fusion reactor systems

International Nuclear Information System (INIS)

Hovingh, J.

1979-01-01

The transfer of energy produced by the interaction of the intense pulses of short-ranged fusion microexplosion products with materials is one of the most difficult problems in inertially-confined fusion (ICF) reactor design. The short time and deposition distance for the energy results in local peak power densities on the order of 10 18 watts/m 3 . High local power densities may cause change of state or spall in the reactor materials. This will limit the structure lifetimes for ICF reactors of economic physical sizes, increasing operating costs including structure replacement and radioactive waste management. Four basic first wall protection methods have evolved: a dry-wall, a wet-wall, a magnetically shielded wall, and a fluid wall. These approaches are distinguished by the way the reactor wall interfaces with fusion debris as well as the way the ambient cavity conditions modify the fusion energy forms and spectra at the first wall. Each of these approaches requires different heat transfer considerations

Inherent/passive safety in fusion power plants

International Nuclear Information System (INIS)

Piet, S.J.; Crocker, J.G.

1986-01-01

The concept of inherent or passive safety for fusion energy is explored, defined, and partially quantified. Four levels of safety assurance are defined, which range from true inherent safety to passive safety to protection via active engineered safeguard systems. Fusion has the clear potential for achieving inherent or passive safety, which should be an objective of fusion research and design. Proper material choice might lead to both inherent/passive safety and high mass power density, improving both safety and economics. When inherent or passive safety is accomplished, fusion will be well on the way to achieving its ultimate potential and to be a truly superior energy source for the future

Fusion reactor design and technology 1986. V. 1

International Nuclear Information System (INIS)

1987-01-01

The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

1st International School of Fusion Reactor Technology "Ettore Majorana"

CERN Document Server

Knoepfel, Heinz; Safety, Environmental Impact and Economic Prospects of Nuclear Fusion

1990-01-01

This book contains the lectures and the concluding discussion of the "Seminar on Safety, Environmental Impact, and Economic Prospects of Nuclear Fusion", which was held at Erice, August 6-12, 1989. In selecting the contributions to this 9th meeting held by the International School of Fusion Reactor Technology at the E. Majorana Center for Scientific Cul­ ture in Erice, we tried to provide a comprehensive coverage of the many interre­ lated and interdisciplinary aspects of what ultimately turns out to be the global acceptance criteria of our society with respect to controlled nuclear fusion. Consequently, this edited collection of the papers presented should provide an overview of these issues. We thus hope that this book, with its extensive subject index, will also be of interest and help to nonfusion specialists and, in general, to those who from curiosity or by assignment are required to be informed on these as­ pects of fusion energy.

Helium cooling of fusion reactors

International Nuclear Information System (INIS)

Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

1994-01-01

On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

Neutronic study of fusion reactor blanket

International Nuclear Information System (INIS)

Barre, F.

1984-02-01

The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

Neutronic study of fusion reactor blanket

International Nuclear Information System (INIS)

Barre, F.

1983-06-01

The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data [fr

A system dynamics model for tritium cycle of pulsed fusion reactor

International Nuclear Information System (INIS)

Zhu, Zuolong; Nie, Baojie; Chen, Dehong

2017-01-01

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

A system dynamics model for tritium cycle of pulsed fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Zhu, Zuolong; Nie, Baojie [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Chen, Dehong, E-mail: dehong.chen@fds.org.cn [Key Laboratory of Neutronics and Radiation Safety, Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China)

2017-05-15

As great challenges and uncertainty exist in achieving steady plasma burning, pulsed plasma burning may be a potential scenario for fusion engineering test reactor, even for fusion DEMOnstration reactor. In order to analyze dynamic tritium inventory and tritium self-sufficiency for pulsed fusion systems, a system dynamics model of tritium cycle was developed on the basis of earlier version of Tritium Analysis program for fusion System (TAS). The model was verified with TRIMO, which was developed by KIT in Germany. Tritium self-sufficiency and dynamic tritium inventory assessment were performed for a typical fusion engineering test reactor. The verification results show that the system dynamics model can be used for tritium cycle analysis of pulsed fusion reactor with sufficient reliability. The assessment results of tritium self-sufficiency indicate that the fusion reactor might only need several hundred gram tritium to startup if achieved high efficient tritium handling ability (Referred ITER: 1 h). And the initial tritium startup inventory in pulsed fusion reactor is determined by the combined influence of pulse length, burn availability, and tritium recycle time. Meanwhile, tritium self-sufficiency can be achieved under the defined condition.

Cost assessment of a generic magnetic fusion reactor

International Nuclear Information System (INIS)

Sheffield, J.; Dory, R.A.; Cohn, S.M.; Delene, J.G.; Parsly, L.F.; Ashby, D.E.T.F.; Reiersen, W.T.

1986-03-01

A generic reactor model is used to examine the economic viability of generating electricity by magnetic fusion. The simple model uses components that are representative of those used in previous reactor studies of deuterium-tritium-burning tokamaks, stellarators, bumpy tori, reversed-field pinches (RFPs), and tandem mirrors. Conservative costing assumptions are made. The generic reactor is not a tokamak; rather, it is intended to emphasize what is common to all magnetic fusion rectors. The reactor uses a superconducting toroidal coil set to produce the dominant magnetic field. To this extent, it is not as good an approximation to systems such as the RFP in which the main field is produced by a plasma current. The main output of the study is the cost of electricity as a function of the weight and size of the fusion core - blanket, shield, structure, and coils. The model shows that a 1200-MW(e) power plant with a fusion core weight of about 10,000 tonnes should be competitive in the future with fission and fossil plants. Studies of the sensitivity of the model to variations in the assumptions show that this result is not sensitively dependent on any given assumption. Of particular importance is the result that a fusion reactor of this scale may be realized with only moderate advances in physics and technology capabilities

What we miss in order to be able to design and build a commercially viable fusion reactor

International Nuclear Information System (INIS)

Andreani, R.

1999-01-01

The paper considers in a critical way the different areas in which work is required to provide sufficient information in view of designing a reliable and attractive fusion reactor. Four main areas of activity are considered: physics, technology, engineering, safety. In physics the trend is positive towards a better understanding of suitable plasma regimes to be confirmed through further experimentation on the operating machines. Engineering has already proven itself by the design and construction of a number of experimental machines. In addition a large data base obtained from design and operation of fission reactors is available. Safety is reaching very satisfactory results in the analysis of the impact of fusion on man and the environment. Where it is still a large unsolved problem is concerning materials capable of standing the harsh fusion environment for an adequate number of years. An intense neutron source is needed in order to allow the necessary developments [it

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Mirror hybrid (fusion--fission) reactor

International Nuclear Information System (INIS)

Bender, D.J.; Lee, J.D.; Neef, W.S.; Devoto, R.S.; Galloway, T.R.; Fink, J.H.; Schultz, K.R.; Culver, D.; Rao, S.

1977-10-01

The reference mirror hybrid reactor design performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. The resulting reactor may thus be viewed as a comparatively near-term goal of the fusion program, and we project improved performance for the hybrid in the future as more advanced technology becomes available

Burning nuclear wastes in fusion reactors

International Nuclear Information System (INIS)

Meldner, H.W.; Howard, W.M.

1979-01-01

A study was made up of actinide burn-up in ICF reactor pellets; i.e. 14 Mev neutron fission of the very long-lived actinides that pose storage problems. A major advantage of pellet fuel region burn-up is safety: only milligrams of highly toxic and active material need to be present in the fusion chamber, whereas blanket burn-up requires the continued presence of tons of actinides in a small volume. The actinide data tables required for Monte Carlo calculations of the burn-up of /sup 241/Am and /sup 243/Am are discussed in connection with a study of the sensitivity to cross section uncertainties. More accurate and complete cross sections are required for realistic quantitative calculations. 13 refs

SOLASE: a conceptual laser fusion reactor design

International Nuclear Information System (INIS)

Conn, R.W.; Abdel-Khalik, S.I.; Moses, G.A.

1977-12-01

The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion ractors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket design, thermal hydraulics, tritium analysis, neutronics calculations, radiation effects, stress analysis, shield design, reactor and plant building layout, maintenance procedures, and power cycle design. The reactor is designed as a 1000 MW/sub e/ unit for central station electric power generation

Past, present and future of the fusion reactors

International Nuclear Information System (INIS)

Rosenbaum P, M.

1992-01-01

Among the alternate technologies that have acquired a special interest in the present decade, we find the nuclear fusion. Within this, the fusion reactors by magnetic confinement of the Tokamak type have shown an increasing technological progress during this period. For this reason, a new strategy, coordinated at international level, has been implemented for the specific development of the nuclear fusion reactors, aimed to face those scientific and technological aspects which still remain, and which will determine their future economic feasibility. (Author)

Safety of Research Reactors. Safety Requirements

International Nuclear Information System (INIS)

2010-01-01

The main objective of this Safety Requirements publication is to provide a basis for safety and a basis for safety assessment for all stages in the lifetime of a research reactor. Another objective is to establish requirements on aspects relating to regulatory control, the management of safety, site evaluation, design, operation and decommissioning. Technical and administrative requirements for the safety of research reactors are established in accordance with these objectives. This Safety Requirements publication is intended for use by organizations engaged in the site evaluation, design, manufacturing, construction, operation and decommissioning of research reactors as well as by regulatory bodies

Safety and deterministic failure analyses in high-beta D-D tokamak reactors

International Nuclear Information System (INIS)

Selcow, E.C.

1984-01-01

Safety and deterministic failure analyses were performed to compare major component failure characteristics for different high-beta D-D tokamak reactors. The primary focus was on evaluating damage to the reactor facility. The analyses also considered potential hazards to the general public and operational personnel. Parametric designs of high-beta D-D tokamak reactors were developed, using WILDCAT as the reference. The size, and toroidal field strength were reduced, and the fusion power increased in an independent manner. These changes were expected to improve the economics of D-D tokamaks. Issues examined using these designs were radiation induced failurs, radiation safety, first wall failure from plasma disruptions, and toroidal field magnet coil failure

Advanced spheromak fusion reactor

International Nuclear Information System (INIS)

Fowler, T.K.

1996-01-01

The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research

Conceptual design of the SlimCS fusion DEMO reactor

International Nuclear Information System (INIS)

Tobita, Kenji; Nishio, Satoshi; Enoeda, Mikio; Nakamura, Hirofumi; Hayashi, Takumi; Asakura, Nobuyuki; Utoh, Hiroyasu; Tanigawa, Hiroyasu; Nishitani, Takeo; Isono, Takaaki; Sakurai, Shinji; Kurita, Genichi; Hayashi, Takao; Oyama, Naoyuki; Liu Changle; Hamamatsu, Kiyotaka; Inoue, Takashi; Ozeki, Takahisa; Sato, Masayasu; Suzuki, Satoshi; Kawashima, Hisato; Ezato, Koichiro; Tsuru, Daigo; Koizumi, Norikiyo; Sakamoto, Keiji; Ando, Masami; Sakamoto, Yoshiteru; Shibama, Yusuke; Suzuki, Takahiro; Takechi, Manabu; Takahashi, Koji; Hirose, Takanori; Sato, Satoru; Nozawa, Takashi; Tanigawa, Hisashi; Kakudate, Satoshi; Kawamura, Yoshinori; Yamanishi, Toshihiko; Hoshino, Tsuyoshi; Ochiai, Kentaro; Ide, Shunsuke; Aiba, Nobuyuki; Shimizu, Katsuhiro; Honda, Mitsuru; Nakamichi, Masaru; Nishi, Hiroshi; Seki, Yoji; Nakamura, Yukiharu; Tsuchiya, Kunihiko; Yoshida, Tohru; Song Yuntao

2010-08-01

This report describes the results of the conceptual design study of the SlimCS fusion DEMO reactor aiming at demonstrating fusion power production in a plant scale and allowing to assess the economic prospects of a fusion power plant. The design study has focused on a compact and low aspect ratio tokamak reactor concept with a reduced-sized central solenoid, which is novel compared with previous tokamak reactor concept such as SSTR (Steady State Tokamak Reactor). Owing to low aspect ratio, the reactor will be capable of having comparatively high beta limit and high elongation (which can elevate the Greenwald density limit), having potential for high power density. The reactor has the main parameters of a major radius of 5.5 m, aspect ratio of 2.6, elongation of 2.0, normalized beta of 4.3, fusion out put of 2.95 GW and average neutron wall load of 3 MW/m 2 . This report covers various aspects of design study including systematic design, physics design, torus configuration, blanket, superconducting magnet, maintenance and building, which were carried out increase the engineering feasibility of the concept. (author)

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.)

Plutonium-239 production rate study using a typical fusion reactor

International Nuclear Information System (INIS)

Faghihi, F.; Havasi, H.; Amin-Mozafari, M.

2008-01-01

The purpose of the present paper is to compute fissile 239 Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m 2 ) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate 239 Pu production rate. Produced 239 Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, 239 Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type

Structural mechanics and reactor safety

International Nuclear Information System (INIS)

Brandes, K.

1983-01-01

Operational safety and reliability of nuclear power plants widely depend on the mechanical behaviour of their structural components and their resistance to the various and complex influences. Durability and consistency of structural components are determined by the kind of strain - during the life - and by environmental conditions. The Conferences on Structural Mechanics in Reactor Technology (SMiRT) are dedicated to the discussion of such questions. The 7th of these Conferences taking place in 2-year increments was held in Chicago in August 1983. The number of contributions again increased, the number of participants slightly decreased. There are some trends in this field worth mentioning, in particular the fact that experience from design and operation of nuclear power plants now available is more and more made use of, and that more and more attention is given the problems of fusion reactors. (orig./HP) [de

Safety considerations in the design of the fusion engineering device

International Nuclear Information System (INIS)

Barrett, R.J.

1983-01-01

Safety considerations play a significant role in the design of a near-term Fusion Engineering Device (FED). For the safety of the general public and the plant workers, the radiation environment caused by the reacting plasma and the potential release of tritium fuel are the dominant considerations. The U.S. Department of Energy (DOE) regulations and guidelines for radiation protection have been reviewed and are being applied to the device design. Direct radiation protection is provided by the device shield and the reactor building walls. Radiation from the activated device components and the tritium fuel is to be controlled with shielding, contamination control, and ventilation. The potential release of tritium from the plant has influenced the selection of reactor building and plant designs and specifications. The safety of the plant workers is affected primarily by the radiation from the activated device components and from plasma chamber debris. The highly activated device components make it necessary to design many of the maintenance activities in the reactor building for totally remote operation. The hot cell facility has evolved as a totally remote maintenance facility due to the high radiation levels of the device components. Safety considerations have had substantial impacts on the design of FED. Several examples of safety-related design impacts are discussed in the paper. Feasible solutions have been identified for all outstanding safety-related items, and additional optimization of these solutions is anticipated in future design studies

Present activities for the preparation of a Japanese draft of structural design guidelines for the experimental fusion reactor

International Nuclear Information System (INIS)

Miya, K.; Muto, Y.; Takatsu, H.; Hada, K.; Koizumi, K.; Jitsukawa, S.; Arai, T.; Ohkawa, Y.; Shimakawa, T.; Aoto, K.; Shiraishi, H.; Takagi, T.; Miki, N.; Takahashi, S.; Sato, K.; Takemasa, F.; Kasaba, M.; Kudough, F.; Fujita, J.; Kajiura, S.; Kinoshita, S.

1996-01-01

Since November 1990, systematic research has been carried out in preparation for a Japanese draft of structural design guidelines for the experimental fusion reactor. This report summarizes the major results of the work and the status of these efforts. A classification of components and definition of operating conditions are proposed on the basis of the ITER-CDA design, in the light of the safety characteristics of the fusion reactor and relevant conventions for the existing fission reactor design code. Specific issues regarding the structural design of the experimental fusion reactor are discussed based on the experimental and analytical work. The validity of the existing structural design method is confirmed for the use of irradiated 316 SS, irrespective of the significant reduction in uniform elongation capability caused by heavy neutron irradiation. Further important phenomena are treated such as magnetic damping, magnetic stiffness and fracture due to electromagnetic forces. Finally, the issues concerned with welding and non-destructive examinations are discussed with relevance to component classification. (orig.)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Radioactive waste management and disposal scenario for fusion power reactors

Energy Technology Data Exchange (ETDEWEB)

Tabara, Takashi; Yamano, Naoki [Sumitomo Atomic Energy Industries Ltd., Tokyo (Japan); Seki, Yasushi; Aoki, Isao

1997-10-01

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a light water reactor (LWR) have been evaluated and compared. At first, the amount and the radioactive level of the radwaste generated in five fusion reactors ware evaluated by an activation calculation code. Next, a possible radwaste disposal scenario applicable to fusion radwaste in Japan is considered and the disposal cost evaluated under certain assumptions. The exposure doses are evaluated for the skyshine of gamma-rays during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical LWR was estimated based on a literature survey and the disposal cost was evaluated using the same assumptions as for the fusion reactors. It is found that the relative cost of disposal is strongly dependent on the cost for interim storage of medium level waste of fusion reactors and the cost of high level waste for the LWR. (author)

Fusion reactor design: On the road to commercialization

International Nuclear Information System (INIS)

Kulcinski, G.L.

1984-01-01

The worldwide effort in fusion is now approximately 2 billion dollars per year and over 12 billion dollars has been invested since 1951 in developing this energy source for the 21st century. A vital component of the past efforts in fusion research has been the conceptual design activities performed by scientists and engineers around the world. Almost 80 such major designs of Tokamak, Mirror, Laser and Ion Beam Reactors have been published and this article discusses how recent conceptual designs have afftected our perception of future fusion reactor performance. (orig.) [de

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

First preliminary design of an experimental fusion reactor

International Nuclear Information System (INIS)

1977-09-01

A preliminary design of a tokamak experimental fusion reactor to be built in the near future is under way. The goals of the reactor are to achieve reactor-level plasma conditions for a sufficiently long operation period and to obtain design, construction and operational experience for the main components of full-scale power reactors. This design covers overall reactor system including plasma characteristics, reactor structure, blanket neutronics, shielding, superconducting magnets, neutral beam injector, electric power supply system, fuel circulating system, reactor cooling system, tritium recovery system and maintenance scheme. The main design parameters are as follows: the reactor fusion power 100 MW, torus radius 6.75 m, plasma radius 1.5 m, first wall radius 1.75 m, toroidal magnet field on axis 6 T, blanket fertile material Li 2 O, coolant He, structural material 316SS and tritium breeding ratio 0.9. (auth.)

The safety designs for the TITAN reversed-field pinch reactor study

International Nuclear Information System (INIS)

Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.; Hoot, C.G.; Schultz, K.R.; Grotz, S.P.; Blanchard, J.; Sharafat, S.; Najmabadi, F.

1989-01-01

TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2 . The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections, and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated at a level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated at a level 2 of safety assurance. (orig.)

The safety designs for the TITAN reversed-field pinch reactor study

International Nuclear Information System (INIS)

Wong, C.P.C.; Cheng, E.T.; Creedon, R.L.; Hoot, C.G.; Schultz, K.R.; Grotz, S.P.; Blanchard, J.P.; Sharafat, S.; Najmabadi, F.

1988-01-01

TITAN is a study to investigate the potential of the reversed-field pinch concept as a compact, high-power density energy system. Two reactor concepts were developed, a self-cooled lithium design with vanadium structure and an aqueous solution loop-in-pool design, both operating at 18 MW/m 2 . The key safety features of the TITAN-I lithium-vanadium blanket design are in material selection, fusion power core configuration selection, lithium piping connections and passive lithium drain tank system. Based on these safety features and results from accident evaluation, TITAN-I can at least be rated as level 3 of safety assurance. For the TITAN-II aqueous loop-in-pool design, the key passive feature is the complete submersion of the fusion power core and the corresponding primary coolant loop system into a pool of low temperature water. Based on this key safety design feature, the TITAN-II design can be rated as level 2 of safety assurance. 7 refs., 2 figs

Summary: Fusion technology, safety and environmental aspects

International Nuclear Information System (INIS)

Matsuda, S.

2003-01-01

The year 2002 was in the middle of successive governmental negotiation toward the start of the ITER Construction. The ITER Engineering Design Activities (EDA) continued until July 2001, and most of the highlighted topics were already reported at the last IAEA Fusion Energy Conference in Montreal or in other opportunities. However, the ITER EDA was followed by the Coordinated Technical Activities that provided a lot of qualitative achievements such as, the search for predictions on operation capabilities based on various data bases and analysis, optimization of the design based on its validating technology R and D. As a consequence, at this conference, major contribution in the field of Fusion Technology was again from ITER, and its related topics occupied about 38% of the total number of contributions of 86. In ITER, physics analysis, predictions and heating/current drive technologies are highlighted. Another key feature at this conference was the progress of study toward steady-state operation in both physics and technology research as well as their application to toroidal devices. Several tokamaks and helical devises are under construction or under design, and most of them incorporate super-conducting magnet for their coils. Studies were made for various types of fusion reactors including Spherical Torus, Tokamaks, Helical systems etc., and their common understandings are progressing through their comparative study. Looking in the near term, but beyond ITER, about 20% of the papers were devoted to the fusion materials and blanket development, with the neutron irradiation facilities for the research. Because of the importance of this field to be implemented in parallel with ITER, more contributions would be expected in future. With these themes in mind, the remaining sections of this paper are arranged in the order of 2) ITER, 3) Toroidal Devices under Construction or under Design, 4) Reactor Technology, 5) Safety and Environment, and 6) Conclusion

Fusion Reactor Materials

International Nuclear Information System (INIS)

Moons, F.

1998-01-01

SCK-CEN's programme on fusion reactor materials includes studies (1) to investigate fracture mechanics of neutron-irradiated beryllium; (2) to describe the helium behaviour in irradiated beryllium at atomic scale; (3) to define the kinetics of beryllium reacting with air or steam; (3) to perform a feasibility study for the testing of integrated blanket modules under neutron irradiation. Progress and achievements in 1997 are reported

Electromagnetic analysis for fusion reactors: status and needs

International Nuclear Information System (INIS)

Turner, L.R.

1983-01-01

Electromagnetic effects have far-reaching implications for the design, operation, and maintenance of future fusion reactors. Two-dimensional (2-D) eddy current computer codes are available, but are of limited value in analyzing reactors. Three-dimensional (3-D) codes are needed, but are only beginning to be developed. Both 2-D and 3-D codes need verification against experimental data, such as that provided by the upcoming FELIX experiments. Coupling between eddy currents and deflections has application in fusion reactor design and is being studied both by analysis and experiment

Fusion reactor remote maintenance study. Final report

International Nuclear Information System (INIS)

Sniderman, M.

1979-04-01

An analysis of a major maintenance operation, the remote replacement of a modular sector of a tokamak reactor, was performed in substantial detail. Specific assumptions were developed which included concepts from various existing designs so that the operation which was studied includes some design features generic to any fusion reactor design. Based on the work performed in this study, the principal conclusions are: (1) It appears feasible to design a tokamak fusion reactor plant with availability comparable to existing fossil and fission plants, but this will require diligence and comprehensive planning during the complete design phase. (2) Since the total fusion program is paced by the success of each device, maintenance considerations must be incorporated into each device during design, even if the device is an experimental unit. (3) Innovative approaches, such as automatic computer controlled operations, should be developed so that large step reductions in planned maintenance times can be achieved

Remote assembly and maintenance of fusion reactors

International Nuclear Information System (INIS)

Becquet, M.C.; Farfaletti-Casali, F.

1991-01-01

This paper intend to present the state of the art in the field of remote assembly and maintenance, including system analysis design and operation for controlled fusion device such as JET, and the next NET and ITER reactors. The operational constraints of fusion reactors with respect to temperature, radiations dose rates and cumulated doses are considered with the resulting design requirements. Concepts like articulated boom, in-vessel vehicle and blanket handling device are presented. The close relations between computer simulations and experimental validation of those concepts are emphasized to ensure reliability of the operational behavior. Mockups and prototypes in reduced and full scale, as operating machines are described to illustrate the progress in remote operations for fusion reactors. The developments achieved at the Institute for System Engineering and Informatics of the Joint Research Center, in the field of remote blanket maintenance, reliability assessment of RH systems and remote cut and welding of lips joints are considered. (author)

Fusion--fission hybrid concepts for laser-induced fusion

International Nuclear Information System (INIS)

Maniscalco, J.

1976-01-01

Fusion-fission hybrid concepts are viewed as subcritical fission reactors driven and controlled by high-energy neutrons from a laser-induced fusion reactor. Blanket designs encompassing a substantial portion of the spectrum of different fission reactor technologies are analyzed and compared by calculating their fissile-breeding and fusion-energy-multiplying characteristics. With a large number of different fission technologies to choose from, it is essential to identify more promising hybrid concepts that can then be subjected to in-depth studies that treat the engineering safety, and economic requirements as well as the neutronic aspects. In the course of neutronically analyzing and comparing several fission blanket concepts, this work has demonstrated that fusion-fission hybrids can be designed to meet a broad spectrum of fissile-breeding and fusion-energy-multiplying requirements. The neutronic results should prove to be extremely useful in formulating the technical scope of future studies concerned with evaluating the technical and economic feasibility of hybrid concepts for laser-induced fusion

Computer simulation of multi-elemental fusion reactor materials

International Nuclear Information System (INIS)

Voertler, K.

2011-01-01

Thermonuclear fusion is a sustainable energy solution, in which energy is produced using similar processes as in the sun. In this technology hydrogen isotopes are fused to gain energy and consequently to produce electricity. In a fusion reactor hydrogen isotopes are confined by magnetic fields as ionized gas, the plasma. Since the core plasma is millions of degrees hot, there are special needs for the plasma-facing materials. Moreover, in the plasma the fusion of hydrogen isotopes leads to the production of high energetic neutrons which sets demanding abilities for the structural materials of the reactor. This thesis investigates the irradiation response of materials to be used in future fusion reactors. Interactions of the plasma with the reactor wall leads to the removal of surface atoms, migration of them, and formation of co-deposited layers such as tungsten carbide. Sputtering of tungsten carbide and deuterium trapping in tungsten carbide was investigated in this thesis. As the second topic the primary interaction of the neutrons in the structural material steel was examined. As model materials for steel iron chromium and iron nickel were used. This study was performed theoretically by the means of computer simulations on the atomic level. In contrast to previous studies in the field, in which simulations were limited to pure elements, in this work more complex materials were used, i.e. they were multi-elemental including two or more atom species. The results of this thesis are in the microscale. One of the results is a catalogue of atom species, which were removed from tungsten carbide by the plasma. Another result is e.g. the atomic distributions of defects in iron chromium caused by the energetic neutrons. These microscopic results are used in data bases for multiscale modelling of fusion reactor materials, which has the aim to explain the macroscopic degradation in the materials. This thesis is therefore a relevant contribution to investigate the

Fusion technology programme

International Nuclear Information System (INIS)

Finken, D.

1985-10-01

KfK is involved in the European Fusion Programme predominantly in the NET and Fusion Technology part. The following fields of activity are covered: Studies for NET, alternative confinement concepts, and needs and issues of integral testing. Research on structural materials. Development of superconducting magnets. Gyrotron development (part of the Physics Programme). Nuclear technology (breeding materials, blanket design, tritium technology, safety and environmental aspects of fusion, remote maintenance). Reported here are status and results of work under contracts with the CEC within the NET and Technology Programme. The aim of the major part of this R and D work is the support of NET, some areas (e.g. materials, safety and environmental impact, blanket design) have a wider scope and address problems of a demonstration reactor. In the current working period, several new proposals have been elaborated to be implemented into the 85/89 Euratom Fusion Programme. New KfK contributions relate to materials research (dual beam and fast reactor irradiations, ferritic steels), to blanket engineering (MHD-effects) and to safety studies (e.g. magnet safety). (orig./GG)

Nuclear design of a very-low-activation fusion reactor

International Nuclear Information System (INIS)

Cheng, E.T.; Hopkins, G.R.

1983-06-01

An investigation was conducted to study the nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE tokamak reactor design

Plutonium-239 production rate study using a typical fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Faghihi, F. [Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Havasi, H.; Amin-Mozafari, M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51154 Shiraz (Iran, Islamic Republic of)

2008-05-15

The purpose of the present paper is to compute fissile {sup 239}Pu material by supposed typical fusion reactor operation to make the fuel requirement for other purposes (e.g. MOX fissile fuel, etc.). It is assumed that there is a fusion reactor has a cylindrical geometry and uses uniformly distributed deuterium-tritium as fuel so that neutron wall load is taken at 10(MW)/(m{sup 2}) . Moreover, the reactor core is surrounded by six suggested blankets to make best performance of the physical conditions described herein. We determined neutron flux in each considered blanket as well as tritium self-sufficiency using two groups neutron energy and then computation is followed by the MCNP-4C code. Finally, material depletion according to a set of dynamical coupled differential equations is solved to estimate {sup 239}Pu production rate. Produced {sup 239}Pu is compared with two typical fission reactors to find performance of plutonium breeding ratio in the case of the fusion reactor. We found that 0.92% of initial U is converted into fissile Pu by our suggested fusion reactor with thermal power of 3000 MW. For comparison, {sup 239}Pu yield of suggested large scale PWR is about 0.65% and for LMFBR is close to 1.7%. The results show that the fusion reactor has an acceptable efficiency for Pu production compared with a large scale PWR fission reactor type.

Control of tritium permeation through fusion reactor strucural materials

International Nuclear Information System (INIS)

Maroni, V.A.

1978-01-01

The intention of this paper is to provide a brief synopsis of the status of understanding and technology pertaining to the dissolution and permeation of tritium in fusion reactor materials. The following sections of this paper attempt to develop a simple perspective for understanding the consequences of these phenomena and the nature of the technical methodology being contemplated to control their impact on fusion reactor operation. Considered in order are: (1) the occurrence of tritium in the fusion fuel cycle, (2) a set of tentative criteria to guide the analysis of tritium containment and control strategies, (3) the basic mechanisms by which tritium may be released from a fusion plant, and (4) the methods currently under development to control the permeation-related release mechanisms. To provide background and support for these considerations, existing solubility and permeation data for the hydrogen isotopes are compared and correlated under conditions to be expected in fusion reactor systems

Environmental and economic assessments of magnetic and inertial fusion energy reactors

Science.gov (United States)

Yamazaki, K.; Oishi, T.; Mori, K.

2011-10-01

Global warming due to rapid greenhouse gas (GHG) emissions is one of the present-day crucial problems, and fusion reactors are expected to be abundant electric power generation systems to reduce human GHG emission amounts. To search for an environmental-friendly and economical fusion reactor system, comparative system studies have been done for several magnetic fusion energy reactors, and have been extended to include inertial fusion energy reactors. We clarify new scaling formulae for the cost of electricity and GHG emission rate with respect to key design parameters, which might be helpful in making a strategy for fusion research development. Comparisons with other conventional electric power generation systems are carried out taking into account the introduction of GHG taxes and the application of the carbon dioxide capture and storage system to fossil power generators.

Fusion reactors-high temperature electrolysis (HTE)

International Nuclear Information System (INIS)

Fillo, J.A.

1978-01-01

Results of a study to identify and develop a reference design for synfuel production based on fusion reactors are given. The most promising option for hydrogen production was high-temperature electrolysis (HTE). The main findings of this study are: 1. HTE has the highest potential efficiency for production of synfuels from fusion; a fusion to hydrogen energy efficiency of about 70% appears possible with 1800 0 C HTE units and 60% power cycle efficiency; an efficiency of about 50% possible with 1400 0 C HTE units and 40% power cycle efficiency. 2. Relative to thermochemical or direct decomposition methods HTE technology is in a more advanced state of development, 3. Thermochemical or direct decomposition methods must have lower unit process or capital costs if they are to be more attractive than HTE. 4. While design efforts are required, HTE units offer the potential to be quickly run in reverse as fuel cells to produce electricity for restart of Tokamaks and/or provide spinning reserve for a grid system. 5. Because of the short timescale of the study, no detailed economic evaluation could be carried out.A comparison of costs could be made by employing certain assumptions. For example, if the fusion reactor-electrolyzer capital installation is $400/(KW(T) [$1000/KW(E) equivalent], the H 2 energy production cost for a high efficiency (about 70 %) fusion-HTE system is on the same order of magnitude as a coal based SNG plant based on 1976 dollars. 6. The present reference design indicates that a 2000 MW(th) fusion reactor could produce as much at 364 x 10 6 scf/day of hydrogen which is equivalent in heating value to 20,000 barrels/day of gasoline. This would fuel about 500,000 autos based on average driving patterns. 7. A factor of three reduction in coal feed (tons/day) could be achieved for syngas production if hydrogen from a fusion-HTE system were used to gasify coal, as compared to a conventional syngas plant using coal-derived hydrogen

Light ion driven inertial fusion reactor concepts

International Nuclear Information System (INIS)

Cook, D.L.; Sweeney, M.A.; Buttram, M.T.; Prestwich, K.R.; Moses, G.A.; peterson, R.R.; Lovell, E.G.; Englestad, R.L.

1980-01-01

The possibility of designing fusion reactor systems using intense beams of light ions has been investigated. concepts for beam production, transport, and focusing on target have been analyzed in light of more conservative target performance estimates. Analyses of the major criteria which govern the design of the beam-target-cavity tried indicate the feasibility of designing power systems at the few hundred megawatt (electric) level. This paper discusses light ion fusion reactor (LIFR) concepts and presents an assessment of the design limitations through quantitative examples

Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

International Nuclear Information System (INIS)

Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

2000-01-01

Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-03-01

A conceptual design study (option C) has been carried out for the fusion experimental reactor (FER). In addition to design of the tokamak reactor and associated systems based on the reference design specifications, feasibility of a water-shield reactor concept was examined as a topical study. The design study for the reference tokamak reactor has produced a reactor concept for the FER, along with major R D items for the concept, based on close examinations on thermal design, electromagnetics, neutronics and remote maintenance. Particular efforts have been directed to the area of electromagnetics. Detailed analyses with close simulation models have been performed on PF coil arrangements and configurations, shell effects of the blanket for plasma position unstability, feedback control, and eddy currents during disruptions. The major design specifications are as follows; Peak fusion power 437 MW Major radius 5.5 m Minor radius 1.1 m Plasma elongation 1.5 Plasma current 5.3 MA Toroidal beta 4 % Field on axis 5.7 T (author)

Blankets for fusion reactors : materials and neutronics

International Nuclear Information System (INIS)

Carvalho, S.H. de.

1980-03-01

The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author) [pt

Fusion reactor pumped laser

International Nuclear Information System (INIS)

Jassby, D.L.

1988-01-01

A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

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

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)

Tritium-related materials problems in fusion reactors

International Nuclear Information System (INIS)

Hickman, R.G.

1976-01-01

Pressing materials problems that must be solved before tritium can be used to produce energy economically in fusion reactors are discussed. The following topics are discussed: (1) breeding tritium, (2) recovering bred tritium, (3) containing tritium, (4) fuel recycling, and (5) laser-fusion fueling

Cost assessment of demo fusion reactor with considering maintenance

International Nuclear Information System (INIS)

Hashizume, Hidetoshi; Kitagoh, Kazutoshi

2003-01-01

The purpose of this study is to perform cost assessment of nuclear fusion reactors in order to draw up commercial plants. A fusion reactor may have a complex configuration to achieve high beta value, which leads to low and instable availability when maintenance is taken into account. Therefore, reactor's availability must be evaluated with considering the influence of the configuration complexity. Furthermore the availability has the strong impact on COE (Cost of Electricity), that is, a fusion reactor with low availability will not be accepted as a commercial plant. Therefore, we developed a new method to calculate availabilities with random numbers, in which the complexity of reactor's configuration could become considered. In addition, we considered the reduction of superconducting coil's maintenance time by introducing remountable magnet system because the coil maintenance requires quite long time in the present technology. The results show that the availability becomes relatively large if the short maintenance time of coils could be achieved, for example, by remountable magnetic systems. (author)

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

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

Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

International Nuclear Information System (INIS)

1975-06-01

The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

Materials design data for fusion reactors

International Nuclear Information System (INIS)

Tavassoli, A.A.F.

1998-01-01

Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.)

Materials design data for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Tavassoli, A.A.F. [CEA Commissariat a l`Energie Atomique, Gif sur Yvette (France). CEREM

1998-10-01

Design data needed for fusion reactors are characterized by the diversity of materials and the complexity of loading situations found in these reactors. In addition, advanced fabrication techniques, such as hot isostatic pressing, envisaged for fabrication of single and multilayered in-vessel components, could significantly change the original materials properties for which the current design rules are written. As a result, additional materials properties have had to be generated for fusion reactors and new structural design rules formulated. This paper recalls some of the materials properties data generated for ITER and DEMO, and gives examples of how these are converted into design criteria. In particular, it gives specific examples for the properties of 316LN-IG and modified 9Cr-1Mo steels, and CuCrZr alloy. These include, determination of tension, creep, isochronous, fatigue, and creep-fatigue curves and their analysis and conversion into design limits. (orig.) 19 refs.

Reduced-activation materials for fusion reactors: An overview of the proceedings

International Nuclear Information System (INIS)

Klueh, R.L.; Packan, N.H.; Gelles, D.S.; Okada, M.

1988-01-01

Some of the most serious safety and environmental concerns for future fusion reactors involve induced radioactivity in the first wall and blanket structures. One problem caused by the induced radioactivity in a reactor constructed from the conventional austenitic and ferritic steels presently being considered as structural materials would be the disposal of the highly radioactive structures after their service lifetimes. To simplify the waste-disposal process, ''low-activation'' or ''reduced-activation'' alloys are being developed. The objective for such materials is that they qualify for shallow land burial, as opposed to the much more expensive deep geologic disposal. This paper reviews these classes of materials for this purpose: austenitic stainless steels, ferritic steels, and vanadium alloys

Critical safety issues in the design of fusion machines

International Nuclear Information System (INIS)

Kramer, W.

1991-01-01

In the course of developing fusion machines both general safety considerations and safety assessments for the various components and systems of actual machines increase in number and become more and more coherent. This is particularly true for the NET/ITER projects where safety analysis plays an increasing role for the design of the machine. Since in a D/T tokamak the radiological hazards will be dominant basic radiological safety objectives are discussed. Critical safety issues as identified in particular by the NET/ITER community are reviewed. Subsequently, issues of major concern are considered both for normal operation and for conceivable accidents. The following accidents are considered to be crucial: Loss of cooling in plasma facing components, loss of vacuum, tritium system failure, and magnet system failure. To mitigate accident consequences a confinement concept based on passive features and multiple barriers including detritiation and filtering has to be applied. The reactor building as final barrier needs special attention to cope with both internal and external hazards. (orig.)

Regulatory aspects of fusion power-lessons from fission plants

International Nuclear Information System (INIS)

Natalizio, A.; Brunnader, H.; Sood, S.K.

1993-01-01

Experience from fission reactors has shown the regulatory process for licensing a nuclear facility to be legalistic, lengthy, unpredictable, and costly. This experience also indicates that much of the regulatory debate is focused on safety margins, that is, the smaller the safety margins the bigger the regulatory debate and the greater the amount of proof required to satisfy the regulatory. Such experience suggests that caution and prudence guide the development of a regulatory regime for fusion reactors. Fusion has intrinsic safety and environmental advantages over fission, which should alleviate significantly, or even eliminate, the regulatory problems associated with fission. The absence of a criticality concern and the absence of fission products preclude a Chernobyl type accident from occurring in a fusion reactor. Although in a fusion reactor there are large inventories of radioactive products that can be mobilized, the total quantity is orders of magnitude smaller than in fission power reactors. The bulk of the radioactivity in a fusion reactor is either activation products in steel structures, or tritium fuel supplies safely stored in the form of a metal tritide in storage beds. The quantity of tritium that can be mobilized under accident conditions is much less than ten million curies. This compares very favorably with a fission product inventory greater than ten billion curies in a fission power reactor. Furthermore, in a fission reactor, all of the reactivity is contained in a steel vessel that is pressurized to about 150 atmospheres, whereas in a fusion reactor, the inventory of radioactive material is dispersed in different areas of the plant, such that it is improbable that a single event could give rise to the release of the entire inventory to the environment. With such significant intrinsic safety advantages there is no a priori need to make fusion requirements/regulations more demanding and more stringent than fission

Preliminary neutronics calculation of fusion-fission hybrid reactor breeding spent fuel assembly

International Nuclear Information System (INIS)

Ma Xubo; Chen Yixue; Gao Bin

2013-01-01

The possibility of using the fusion-fission hybrid reactor breeding spent fuel in PWR was preliminarily studied in this paper. According to the fusion-fission hybrid reactor breeding spent fuel characteristics, PWR assembly including fusion-fission hybrid reactor breeding spent fuel was designed. The parameters such as fuel temperature coefficient, moderator temperature coefficient and their variation were investigated. Results show that the neutron properties of uranium-based assembly and hybrid reactor breeding spent fuel assembly are similar. The design of this paper has a smaller uniformity coefficient of power at the same fissile isotope mass percentage. The results will provide technical support for the future fusion-fission hybrid reactor and PWR combined with cycle system. (authors)

Characteristics of irradiation creep in the first wall of a fusion reactor

International Nuclear Information System (INIS)

Coghlan, W.A.; Mansur, L.K.

1981-01-01

A number of significant differences in the irradiation environment of a fusion reactor are expected with respect to the fission reactor irradiation environment. These differences are expected to affect the characteristics of irradiation creep in the fusion reactor. Special conditions of importance are identified as the (1) large number of defects produced per pka, (2) high helium production rate, (3) cyclic operation, (4) unique stress histories, and (5) low temperature operations. Existing experimental data from the fission reactor environment is analyzed to shed light on irradiation creep under fusion conditions. Theoretical considerations are used to deduce additional characteristics of irradiation creep in the fusion reactor environment for which no experimental data are available

Evaluation of the impact of a committed site on fusion reactor development

International Nuclear Information System (INIS)

Reid, R.L.; Nagy, A.

1979-01-01

The technical and economic merits of a committed fusion site for development of tokamak, mirror, and EBT reactor from ignition through demo phases were evaluated. Schedule compression resulting from evolving several reactor concepts and/or phases on a committed site as opposed to sequential use of independent sites was estimated. Land, water, and electrical power requirements for a committed fusion site were determined. A conceptual plot plan for siting three fusion reactors on a committed site was configured. Reactor support equipment common to the various concepts was identified as candidates for sharing. Licensing issues for fusion plants were briefly addressed

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

RB research reactor safety report

International Nuclear Information System (INIS)

Sotic, O.; Pesic, M.; Vranic, S.

1979-04-01

This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

The ITER fusion reactor and its role in the development of a fusion power plant

International Nuclear Information System (INIS)

McLean, A.

2002-01-01

Energy from nuclear fusion is the future source of sustained, full life-cycle environmentally benign, intrinsically safe, base-load power production. The nuclear fusion process powers our sun, innumerable other stars in the sky, and some day, it will power the Earth, its cities and our homes. The International Thermonuclear Experimental Reactor, ITER, represents the next step toward fulfilling that promise. ITER will be a test bed for key steppingstones toward engineering feasibility of a demonstration fusion power plant (DEMO) in a single experimental step. It will establish the physics basis for steady state Tokamak magnetic containment fusion reactors to follow it, exploring ion temperature, plasma density and containment time regimes beyond the breakeven power condition, and culminating in experimental fusion self-ignition. (author)

Extrap conceptual fusion reactor design study

International Nuclear Information System (INIS)

Eninger, J.E; Lehnert, B.

1987-12-01

A study has recently been initiated to asses the fusion reactor potential of the Extrap concept. A reactor model is defined that fulfills certain economic and environmental criteria. This model is applied to Extrap and a reference reactor is outlined. The design is optimized by varying parameters subject to both physics and engineering constraints. Several design options are examined and key engineering issues are identified and addressed. Some preliminary results and conclusions of this work are summarized. (authors)

Materials needs for compact fusion reactors

International Nuclear Information System (INIS)

Krakowski, R.A.

1983-01-01

The economic prospects for magnetic fusion energy can be dramatically improved if for the same total power output the fusion neutron first-wall (FW) loading and the system power density can be increased by factors of 3 to 5 and 10 to 30, respectively. A number of compact fusion reactor embodiments have been proposed, all of which would operate with increased FW loadings, would use thin (0.5 to 0.6 m) blankets, and would confine quasi-steady-state plasma with resistive, water-cooled copper or aluminum coils. Increased system power density (5 to 15 MWt/m 3 versus 0.3 to 0.5 MW/m 3 ), considerably reduced physical size of the fusion power core (FPC), and appreciably reduced economic leverage exerted by the FPC and associated physics result. The unique materials requirements anticipated for these compact reactors are outlined against the well documented backdrop provided by similar needs for the mainline approaches. Surprisingly, no single materials need that is unique to the compact systems is identified; crucial uncertainties for the compact approaches must also be addressed by the mainline approaches, particularly for in-vacuum components (FWs, limiters, divertors, etc.)

RB research reactor Safety Report

International Nuclear Information System (INIS)

Sotic, O.; Pesic, M.; Vranic, S.

1979-04-01

This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

Numerical analysis of magnetoelastic coupled buckling of fusion reactor components

International Nuclear Information System (INIS)

Demachi, K.; Yoshida, Y.; Miya, K.

1994-01-01

For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated

Health physics in fusion reactor design

International Nuclear Information System (INIS)

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

1984-06-01

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

Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

International Nuclear Information System (INIS)

Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

1997-01-01

Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

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)

IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors

Hydrogen production from high temperature electrolysis and fusion reactor

International Nuclear Information System (INIS)

Dang, V.D.; Steinberg, J.F.; Issacs, H.S.; Lazareth, O.; Powell, J.R.; Salzano, F.J.

1978-01-01

Production of hydrogen from high temperature electrolysis of steam coupled with a fusion reactor is studied. The process includes three major components: the fusion reactor, the high temperature electrolyzer and the power conversion cycle each of which is discussed in the paper. Detailed process design and analysis of the system is examined. A parametric study on the effect of process efficiency is presented

Failure rate data for fusion safety and risk assessment

International Nuclear Information System (INIS)

Cadwallader, L.C.

1993-01-01

The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

Utilization of fusion neutrons in the tokamak fusion test reactor for blanket performance testing and other nuclear engineering experiments

International Nuclear Information System (INIS)

Caldwell, C.S.; Pettus, W.G.; Schmotzer, J.K.; Welfare, F.; Womack, R.

1979-01-01

In addition to developing a set of reacting-plasma/blanket-neutronics benchmark data, the TFTR fusion application experiments would provide operational experience with fast-neutron dosimetry and the remote handling of blanket modules in a tokamak reactor environment; neutron streaming and hot-spot information invaluable for the optimal design of penetrations in future fusion reactors; and the identification of the most damage-resistant insulators for a variety of fusion-reactor components

Magnet design considerations for Tokamak fusion reactors

International Nuclear Information System (INIS)

Purcell, J.R.; Chen, W.; Thomas, R.

1976-01-01

Design problems for superconducting ohmic heating and toroidal field coils for large Tokamak fusion reactors are discussed. The necessity for making these coils superconducting is explained, together with the functions of these coils in a Tokamak reactor. Major problem areas include materials related aspects and mechanical design and cryogenic considerations. Projections and comparisons are made based on existing superconducting magnet technology. The mechanical design of large-scale coils, which can contain the severe electromagnetic loading and stress generated in the winding, are emphasized. Additional major tasks include the development of high current conductors for pulsed applications to be used in fabricating the ohmic heating coils. It is important to note, however, that no insurmountable technical barriers are expected in the course of developing superconducting coils for Tokamak fusion reactors. (Auth.)

Optical design considerations for laser fusion reactors

International Nuclear Information System (INIS)

Monsler, M.J.; Maniscalco, J.A.

1977-09-01

The plan for the development of commercial inertial confinement fusion (ICF) power plants is discussed, emphasizing the utilization of the unique features of laser fusion to arrive at conceptual designs for reactors and optical systems which minimize the need for advanced materials and techniques requiring expensive test facilities. A conceptual design for a liquid lithium fall reactor is described which successfully deals with the hostile x-ray and neutron environment and promises to last the 30 year plant lifetime. Schemes for protecting the final focusing optics are described which are both compatible with this reactor system and show promise of surviving a full year in order to minimize costly downtime. Damage mechanisms and protection techniques are discussed, and a recommendation is made for a high f-number metal mirror final focusing system

Thermal energy and bootstrap current in fusion reactor plasmas

International Nuclear Information System (INIS)

Becker, G.

1993-01-01

For DT fusion reactors with prescribed alpha particle heating power P α , plasma volume V and burn temperature i > ∼ 10 keV specific relations for the thermal energy content, bootstrap current, central plasma pressure and other quantities are derived. It is shown that imposing P α and V makes these relations independent of the magnitudes of the density and temperature, i.e. they only depend on P α , V and shape factors or profile parameters. For model density and temperature profiles analytic expressions for these shape factors and for the factor C bs in the bootstrap current formula I bs ∼ C bs (a/R) 1/2 β p I p are given. In the design of next-step devices and fusion reactors, the fusion power is a fixed quantity. Prescription of the alpha particle heating power and plasma volume results in specific relations which can be helpful for interpreting computer simulations and for the design of fusion reactors. (author) 5 refs

Assessment of tritium breeding requirements for fusion power reactors

International Nuclear Information System (INIS)

Jung, J.

1983-12-01

This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios

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

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Nuclear data for fusion reactor technology

International Nuclear Information System (INIS)

1988-06-01

The meeting was organized in four sessions and four working groups devoted to the following topics: Requirements of nuclear data for fusion reactor technology (6 papers); Status of experimental and theoretical investigations of microscopic nuclear data (10 papers); Status of existing libraries for fusion neutronic calculations (5 papers); and Status of integral experiments and benchmark tests (6 papers). A separate abstract was prepared for each of these papers

TORFA - toroidal reactor for fusion applications

International Nuclear Information System (INIS)

Jassby, D.L.

1980-09-01

The near-term goal of the US controlled fusion program should be the development, for practical applications, of an intense, quasi-steady, reliable 14-MeV neutron source with an electrical utilization efficiency at least 10 times larger than the value characterizing beam/solid-target neutron generators. This report outlines a method for implementing that goal, based on tokamak fusion reactors featuring resistive toroidal-field coils designed for ease of demountability

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

The ARIES tokamak fusion reactor study

International Nuclear Information System (INIS)

Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

1989-01-01

The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

MINIMARS: An attractive small tandem mirror fusion reactor

International Nuclear Information System (INIS)

Perkins, L.J.; Logan, B.G.; Doggett, J.N.; Devoto, R.S.

1986-01-01

Through the innovative design of a novel end plug scheme employing octopole MHD stabilization, the authors present the conceptual design of ''MINIMARS'', a small commercial fusion reactor based on the tandem mirror principle. The current baseline for MINIMARS has a net electric output of 600 MWe and they have configured the design for short construction times, factory-built modules, inherently safe blanket systems, and multiplexing in station sizes of ≅ 600-2400 MWe. They demonstrate that the compact octopole end cell provides a number of advantages over the more conventional quadrupole (yin-yang) end cell encountered in the MARS tandem mirror reactor study, and enables ignition to be achieved with much shorter central cell lengths. Accordingly, being economic in small sizes, MINIMARS provides an attractive alternative to the more conventional larger conceptual fusion reactors encountered to date, and would contribute significantly to the lowering of utility financial risk in a developing fusion economy

Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

International Nuclear Information System (INIS)

Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

1996-01-01

Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

Synfuels production from fusion reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.; Steinberg, M.

The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and supplement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Depending on design, electric generation efficiencies of approximately 40 to 60 percent and hydrogen production efficiencies by high temperature electrolysis of approximately 50 to 70 percent are projected for fusion reactors using high temperature blankets

Engineering and physics of high-power-density, compact, reversed-field-pinch fusion reactors

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.; Krakowski, R.A.; Schultz, K.R.; Steiner, D.

1989-01-01

The technical feasibility and key developmental issues of compact, high-power-density Reversed-Field-Pinch (RFP) reactors are the primary results of the TITAN RFP reactor study. Two design approaches emerged, TITAN-I and TITAN-II, both of which are steady-state, DT-burning, circa 1000 MWe power reactors. The TITAN designs are physically compact and have a high neutron wall loading of 18 MW m 2 . Detailed analyses indicate that: a) each design is technically feasible; b) attractive features of compact RFP reactors can be realized without sacrificing the safety and environmental potential of fusion; and c) major features of this particular embodiment of the RFP reactor are retained in a design window of neutron wall loading ranging from 10 to 20 MW/m 2 . A major product of the TITAN study is the identification and quantification of major engineering and physics requirements for this class of RFP reactors. These findings are the focus of this paper. (author). 26 refs.; 4 figs.; 1 tab

Material for fusion reactor

International Nuclear Information System (INIS)

Abhishek, Anuj; Ranjan, Prem

2011-01-01

To make nuclear fusion power a reality, the scientists are working restlessly to find the materials which can confine the power generated by the fusion of two atomic nuclei. A little success in this field has been achieved, though there are still miles to go. Fusion reaction is a special kind of reaction which must occur at very high density and temperature to develop extremely large amount of energy, which is very hard to control and confine within using the present techniques. As a whole it requires the physical condition that rarely exists on the earth to carry out in an efficient manner. As per the growing demand and present scenario of the world energy, scientists are working round the clock to make effective fusion reactions to real. In this paper the work presently going on is considered in this regard. The progress of the Joint European Torus 2010, ITER 2005, HiPER and minor works have been studied to make the paper more object oriented. A detailed study of the technological and material requirement has been discussed in the paper and a possible suggestion is provided to make a contribution in the field of building first ever nuclear fusion reactor

Linear stability of toroidal Alfvén eigenmodes in the Chinese Fusion Engineering Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Yang, Wenjun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); University of Science and Technology of China, Hefei, Anhui 230026 (China); Li, Guoqiang, E-mail: ligq@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Hu, Youjun; Gao, Xiang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

2017-01-15

The Chinese Fusion Engineering Test Reactor (CFETR) is under design. It aims to fill the gaps between ITER and DEMO. In the reactor, the deuterium-tritium fusion reaction and the auxiliary heating will generate a lot of energetic particles. It is possible that these energetic particles will drive toroidal Alfvén eigenmode (TAE) instabilities under the conditions of CFETR plasma parameters. These instabilities can result in energetic particles redistribution or loss, so it’s vital to study TAE instabilities in CFETR. The aim of this paper is to study the possibility of reducing TAE instabilities by changing safety factor profiles in CFETR. NOVA and NOVA-K codes are used to study TAE stability. The equilibria are constructed using the CORSICA code. Safety factor profiles are selected as the three typical profiles of ITER scenarios. For the three different safety factor profiles, we use NOVA to scan and calculate their continuum spectrum and eigenmode structures, then use NOVA-K to calculate the different damping and driving mechanisms for different toroidal mode numbers. The numerical calculations show that if the safety factor profiles are chosen appropriately, then all the TAEs can be stable. Thus, it’s possible to reduce the TAE instabilities by changing safety factor profiles in CFETR. We also scan the temperature and density profiles to see their effects on the TAE instabilities. It shows that the TAE instabilities keep unchanged for a wide range of profiles.

Status of researches in the field of safety of pressurized water reactors

International Nuclear Information System (INIS)

Couturier, Jean; Schwarz, Michel

2017-01-01

This collective publication proposes a synthesis of the status of researches performed in the field of safety of pressurized water reactors. They may discuss past, current and projected research works, involved actors, or lessons learned from these works. The authors propose a presentation of some research tools privileged by the IRSN for these researches: the CABRI and PHEBUS reactors, the GALAXIE experimental platform, and some other installations. Then they address researches related to loss-of-coolant accidents (two-phase thermohydraulics, fuel rod behaviour), to reactivity accidents, to accidents related to dewatering of irradiated fuel storage pools, to fires, to extreme aggressions of natural origin (earthquake, extreme flooding), to core fusion accidents (core heating and fusion within the vessel, vessel failure and apron erosion by corium, containment enclosure dynamic loading, release of radioactive products), to the behaviour of nuclear plant important metallic or civil works components and notably to their ageing, to organisational and human factors or more generally to social and human sciences (design of control rooms, safety organisation and management in EDF nuclear plants), and to other issues and research perspectives

Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

International Nuclear Information System (INIS)

Yu, Zhang-cheng; Xie, Heng

2014-01-01

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs

Recycling fusion materials

International Nuclear Information System (INIS)

Ooms, L.

2005-01-01

The inherent safety and environmental advantages of fusion power in comparison with other energy sources play an important role in the public acceptance. No waste burden for future generations is therefore one of the main arguments to decide for fusion power. The waste issue has thus been studied in several documents and the final conclusion of which it is stated that there is no permanent disposal waste needed if recycling is applied. But recycling of fusion reactor materials is far to be obvious regarding mostly the very high specific activity of the materials to be handled, the types of materials and the presence of tritium. The main objective of research performed by SCK-CEN is to study the possible ways of recycling fusion materials and analyse the challenges of the materials management from fusion reactors, based on current practices used in fission reactors and the requirements for the manufacture of fusion equipment

Fusion reactor horizontal versus vertical maintenance approach

International Nuclear Information System (INIS)

Charruyer, Ph.; Djerassi, H.; Leger, D.; Maupou, M.; Rouillard, J.; Salpietro, E.; Holloway, C.; Suppan, A.

1987-01-01

This paper concerns the comparison of horizontal versus vertical maintenance options of internal components (blanket and segment) of fusion reactors NET (Next European Torus) and INTOR Design. The described mechanical options are taken to ensure the handling of internals with the required precision, taking into account the problems raised by the safety and confinement requirements. Handling is obviously performed remotely. The option comparisons are performed according to the criteria of feasibility, building size, duration of maintenance operations, safety, flexibility, availability and cost. The first conclusions point on that the vertical handling option offers advantages, as regards the ease of handling and confinement possibilities. From the building size point of view, the two solutions are almost equivalent, while other criteria do not provide a basis for choice. It is emphasized that the confinement option C.T.U. (Containment Transfer Unit) or T.I.C. (Tight Intermediate Confinement) should be the major factor in determining the best options. In additions, a cost comparative analysis emphasizes the best cost/benefit ratio for the different options studied

ELMO Bumpy Torus fusion-reactor design study

International Nuclear Information System (INIS)

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

1981-01-01

A complete power plant design of a 1200-MWe ELMO Bumpy Torus Reactor (EBTR) is described that emphasizes those features that are unique to the EBT confinement concept, with subsystems and balance-of-plant items that are generic to magnetic fusion being adopted from past, more extensive tokamak reactor designs

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

1998-01-01

The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

International Nuclear Information System (INIS)

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

2001-01-01

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

Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Nordic studies in reactor safety

International Nuclear Information System (INIS)

Pershagen, N.

1993-01-01

The Nordic Nuclear Safety Research Programme SIK programme in reactor safety is part of a major joint Nordic research effort in nuclear safety. The report summarizes the achievements of the SIK programme, which was carried out during 1990-1993 in collaboration between Nordic nuclear utilities, safety authorities, and research institutes. Three main projects were successfully completed dealing with: 1) development and application of a living PSA concept for monitoring the risk of core damage, and of safety indicators for early warning of possible safety problems; 2) review and intercomparison of severe accident codes, case studies of potential core melt accidents in nordic reactors, development of chemical models for the MAAP code, and outline of a system for computerized accident management support; 3) compilation of information about design and safety features of neighbouring reactors in Germany, Lithuania and Russia, and for naval reactors and nuclear submarines. The report reviews the state-of-the-art in each subject matter as an introduction to the individual project summaries. The main findings of each project are highlighted. The report also contains an overview of reactor safety research in the Nordic countries and a summary of fundamental reactor safety principles. (au) (69 refs.)

Fusion reactor high vacuum pumping

International Nuclear Information System (INIS)

Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M.

1992-01-01

This paper reports on recent experiments which have shown the practicality of using activated carbon (coconut charcoal) at 4K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were satisfactory. The long-term effects of tritium on the charcoal/cement system developed by Grumman and LLNL was now known; therefore a program was undertaken to see what, if any, effect long-term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77 K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately one-third way through, and after the exposure. Modest effects were noted which would not seriously restrict the use of charcoal as a cryosorber for fusion reactor high-vacuum pumping applications

Vanadium-base alloys for fusion reactor applications

International Nuclear Information System (INIS)

Smith, D.L.; Loomis, B.A.; Diercks, D.R.

1984-10-01

Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined

Vanadium-base alloys for fusion reactor applications

Energy Technology Data Exchange (ETDEWEB)

Smith, D.L.; Loomis, B.A.; Diercks, D.R.

1984-10-01

Vanadium-base alloys offer potentially significant advantages over other candidate alloys as a structural material for fusion reactor first wall/blanket applications. Although the data base is more limited than that for the other leading candidate structural materials, viz., austenitic and ferritic steels, vanadium-base alloys exhibit several properties that make them particularly attractive for the fusion reactor environment. This paper presents a review of the structural material requirements, a summary of the materials data base for selected vanadium-base alloys, and a comparison of projected performance characteristics compared to other candidate alloys. Also, critical research and development (R and D) needs are defined.

Progress of electromagnetic analysis for fusion reactors

International Nuclear Information System (INIS)

Takagi, T.; Ruatto, P.; Boccaccini, L.V.

1998-01-01

This paper describes the recent progress of electromagnetic analysis research for fusion reactors including methods, codes, verification tests and some applications. Due to the necessity of the research effort for the structural design of large tokamak devices since the 1970's with the help of the introduction of new numerical methods and the advancement of computer technologies, three-dimensional analysis methods have become as practical as shell approximation methods. The electromagnetic analysis is now applied to the structural design of new fusion reactors. Some more modeling and verification tests are necessary when the codes are applied to new materials with nonlinear material properties. (orig.)

Fusion--fission hybrid reactors based on the laser solenoid

International Nuclear Information System (INIS)

Steinhauer, L.C.; Taussig, R.T.; Quimby, D.C.

1976-01-01

Fusion-fission reactors, based on the laser solenoid concept, can be much smaller in scale than their pure fusion counterparts, with moderate first-wall loading and rapid breeding capabilities (1 to 3 tonnes/yr), and can be designed successfully on the basis of classical plasma transport properties and free-streaming end-loss. Preliminary design information is presented for such systems, including the first wall, pulse coil, blanket, superconductors, laser optics, and power supplies, accounting for the desired reactor performance and other physics and engineering constraints. Self-consistent point designs for first and second generation reactors are discussed which illustrate the reactor size, performance, component parameters, and the level of technological development required

Fusion reactor design and technology program in China

International Nuclear Information System (INIS)

Huang, J.H.

1994-01-01

A fusion-fission hybrid reactor program was launched in 1987. The purpose of development of the hybrid reactor is twofold: to solve the problem of nuclear fuel supply for an expected large-scale development of fission reactor plants, and to maintain the momentum of fusion research. The program is described and the activities and progress of the program are presented. Two conceptual designs of an engineering test reactor with tokamak configuration were developed at the Southwestern Institute of Physics and the Institute of Plasma Physics. The results are a tokamak engineering test breeder (TETB) series design and a fusion-fission hybrid reactor design (SSEHR), characterized by a liquid-Li self-cooled blanket and an He-cooled solid tritium breeder blanket respectively. In parallel with the design studies, relevant technological experiments on a small or medium scale have been supported by this program. These include LHCD, ICRH and pellet injection in the area of plasma engineering; neutronics integral experiments with U, Pu, Fe and Be; various irradiation tests of austenitic and ferritic steels, magnetohydrodynamic (MHD) pressure drop experiments using a liquid metal loop; research into permeation barriers for tritium and hydrogen isotopes; solid tritium breeder tests using an in-situ loop in a fission reactor. All these experiments have proceeded successfully. The second step of this program is now starting. It seems reasonable that most of the research carried out in the first step will continue. ((orig.))

Canada's Fusion Program

International Nuclear Information System (INIS)

Jackson, D. P.

1990-01-01

Canada's fusion strategy is based on developing specialized technologies in well-defined areas and supplying these technologies to international fusion projects. Two areas are specially emphasized in Canada: engineered fusion system technologies, and specific magnetic confinement and materials studies. The Canadian Fusion Fuels Technology Project focuses on the first of these areas. It tritium and fusion reactor fuel systems, remote maintenance and related safety studies. In the second area, the Centre Canadian de fusion magnetique operates the Tokamak de Varennes, the main magnetic fusion device in Canada. Both projects are partnerships linking the Government of Canada, represented by Atomic Energy of Canada Limited, and provincial governments, electrical utilities, universities and industry. Canada's program has extensive international links, through which it collaborates with the major world fusion programs, including participation in the International Thermonuclear Experimental Reactor project

The TITAN reversed-field-pinch fusion reactor study

International Nuclear Information System (INIS)

1990-01-01

This paper on titan plasma engineering contains papers on the following topics: reversed-field pinch as a fusion reactor; parametric systems studies; magnetics; burning-plasma simulations; plasma transient operations; current drive; and physics issues for compact RFP reactors

Waste management for JAERI fusion reactors

International Nuclear Information System (INIS)

Tobita, K.; Nishio, S.; Konishi, S.; Jitsukawa, S.

2004-01-01

In the fusion reactor design study at Japan Atomic Energy Institute (JAERI), several waste management strategies were assessed. The assessed strategies are: (1) reinforced neutron shield to clear the massive ex-shielding components from regulatory control; (2) low aspect ratio tokamak to reduce the total waste; (3) reuse of liquid metal breeding material and neutron shield. Combining these strategies, the weight of disposal waste from a low aspect ratio reactor VECTOR is expected to be comparable with the metal radwaste from a light water reactor (∼4000 t)

ITER: the first experimental fusion reactor

International Nuclear Information System (INIS)

Rebut, P.H.

1995-01-01

The International Thermonuclear Experimental Reactor (ITER) project is a multiphased project, at present proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement between the European Atomic Energy Community, the Government of Japan, the Government of the USA and the Government of Russia (''the parties''). The project is based on the tokamak, a Russian invention which has been brought to a high level of development and progress in all major fusion programs throughout the world.The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production and to test technologies for a demonstration fusion power plant. During the extended performance phase of ITER, it will demonstrate the characteristics of a fusion power plant, producing more than 1500MW of fusion power.The objective of the engineering design activity (EDA) phase is to produce a detailed, complete and fully integrated engineering design of ITER and all technical data necessary for the future decision on the construction of ITER.The ITER device will be a major step from present fusion experiments and will encompass all the major elements required for a fusion reactor. It will also require the development and the implementation of major new components and technologies.The inside surface of the plasma containment chamber will be designed to withstand temperature of up to 500 C, although normal operating temperatures will be substantially lower. Materials will have to be carefully chosen to withstand these temperatures, and a high neutron flux. In addition, other components of the device will be composed of state-of-the-art metal alloys, ceramics and composites, many of which are now in the early stage of development of testing. (orig.)

Hefei experimental hybrid fusion-fission reactor conceptual design

International Nuclear Information System (INIS)

Qiu Lijian; Luan Guishi; Xu Qiang

1992-03-01

A new concept of hybrid reactor is introduced. It uses JET-like(Joint European Tokamak) device worked at sub-breakeven conditions, as a source of high energy neutrons to induce a blanket fission of depleted uranium. The solid breeding material and helium cooling technique are also used. It can produce 100 kg of 239 Pu per year by partial fission suppressed. The energy self-sustained of the fusion core is not necessary. Plasma temperature is maintained by external 20 MW ICRF (ion cyclotron resonance frequency) and 10 MW ECRF (electron cyclotron resonance frequency) heating. A steady state plasma current at 1.5 Ma is driven by 10 MW LHCD (lower hybrid current driven). Plasma density will be kept by pellet injection. ICRF can produce a high energy tail in ion distribution function and lead to significant enhancement of D-T reaction rate by 2 ∼ 5 times so that the neutron source strength reaches to the level of 1 x 10 19 n/s. This system is a passive system. It's power density is 10 W/cm 3 and the wall loading is 0.6 W/cm 2 that is the lower limitation of fusion and fission technology. From the calculation of neutrons it could always be in sub-critical and has intrinsic safety. The radiation damage and neutron flux distribution on the first wall are also analyzed. According to the conceptual design the application of this type hybrid reactor earlier is feasible

Conceptual design study of Fusion Experimental Reactor (FY87FER)

International Nuclear Information System (INIS)

1988-05-01

The design study of Fusion Experimental Reactor(FER) which has been proposed to be the next step fusion device has been conducted by JAERI Reactor System Laboratory since 1982 and by FER design team since 1984. This is the final report of the FER design team program and describes the results obtained in FY1987 (partially in FY1986) activities. The contents of this report consist of the reference design which is based on the guideline in FY1986 by the Subcomitees set up in Nuclear Fusion Council of Atomic Energy Commission of Japan, the Low-Physics-Risk reactor design for achieving physics mission more reliably and the system study of FER design candidates including above two designs. (author)

Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

International Nuclear Information System (INIS)

Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

2016-01-01

Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

Energy Technology Data Exchange (ETDEWEB)

Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

2016-11-01

Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

Bringing fusion electric power closer

International Nuclear Information System (INIS)

Kintner, E.

1977-01-01

A review of the controlled fusion research program is given. The tokamak research program is described. Beam injection heating, control systems, and the safety of fusion reactors are topics that are also discussed

Challenges of designing fusion reactors for remote maintainability

International Nuclear Information System (INIS)

Masson, L.S.

1981-01-01

One of the major problems faced by the fusion community is the development of the high level of reliability required to assure that fusion will be a viable commercial power source. Much of the responsibility for solving this problem falls directly on the designer in developing concepts that have a high level of maintainability for the next generation engineering oriented reactors; and long range, in developing full maintainability for the more complicated commercial concepts with their required high level of on-line time. The near-term challenge will include development of unique design concepts to perform inspection, maintenance, replacement, and testing under the stringent conditions imposed by the next generation engineering oriented machines. The long range challenge will focus on basic design concepts that will enable the full maintainability required by commercial fusion. In addition to the purely technical challenges, the fusion community is also faced with the problem of developing programmatic means to assure that reactor maintenance issues are given proper and timely emphasis as the nuclear phase of fusion is approached

Safety of Research Reactors. Specific Safety Requirements (French Edition)

International Nuclear Information System (INIS)

2017-01-01

This Safety Requirements publication establishes requirements for all main areas of safety for research reactors, with particular emphasis on requirements for design and operation. It explains the safety objectives and concepts that form the basis for safety and safety assessment for all stages in the lifetime of a research reactor. Technical and administrative requirements for the safety of new research reactors are established in accordance with these objectives and concepts, and they are to be applied to the extent practicable for existing research reactors. The safety requirements established in this publication for the management of safety and regulatory supervision apply to site evaluation, design, manufacturing, construction, commissioning, operation (including utilization and modification), and planning for decommissioning of research reactors (including critical assemblies and subcritical assemblies). The publication is intended for use by regulatory bodies and other organizations with responsibilities in these areas and in safety analysis, verification and review, and the provision of technical support.

Study on Fusion Safety Infrastructure using ISAM

International Nuclear Information System (INIS)

Oh, Kyemin; Kang, Myungsuk; Heo, Gyunyoung; Kim, Hyoungchan

2013-01-01

The regulation of nuclear facilities have checked and managed safety throughout the entire process from design, construction, operation and decommissioning. Also, the same meaning as the regulatory requirements and design requirements, it will be important indicators for detailed design of K-DEMO. K-DEMO has many uncertainties because it is in conceptual design phase. Also, there is no reference material because demonstration scale fusion power plants were not operated yet in overseas. So, hazard that threaten the integrity of K-DEMO have to be defined preferentially to define regulatory or design requirements. This study proposed method that educe regulatory or design requirements and introduce web-based cloud infrastructure to perform renewal and sharing of information related with safety that is required in the study rapidly as a part of the R and D program funded by National Fusion Research Institute of Korea (NFRI). We have been performing QSR and PIRT in accordance with development of fusion DEMO plant, and preparing OPT, PSA and DPA for regulation requirements. This study introduces our recent research activities about ISAM for fusion and CCI built for expert and extant safety related information. Unlike fission, nuclear fusion's safety goal is non-evacuation of the public during an accident. To satisfy this goal not only various safety issues should be analyzed, but safety objectives, regulatory requirements, and design variables should also be established in detailed design phase. The web-based cloud infrastructure proposed in this paper will be able to offer input data of future studies and, it is expected to contribute on general and technical safety principles for national fusion power plant technology plan

Prospects for alternative Fusion Fuels

International Nuclear Information System (INIS)

Glancy, J.

1986-01-01

The author has worked on three different magnetic confinement concepts for alternate fusion fueled reactors: tokamaks; tanden mirrors, and reversed field pinches. The focus of this article is on prospects for alternate fusion fuels as the author sees them relative to the other choices: increased numbers of coal plants, fission reactors, renewables, and D-T fusion. Discussion is limited on the consideration of alternate fusion fuels to the catalyzed deuterium-deuterium fuel cycle. Reasons for seeking an alternate energy source are cost, a more secure fuel supply, environmental impact and safety. The technical risks associated with development of fusion are examined briefly

Transmutation of actinide 237Np with a fusion reactor and a hybrid reactor

International Nuclear Information System (INIS)

Feng, K.M.; Huang, J.H.

1994-01-01

The use of fusion reactors to transmute fission reactor wastes to stable species is an attractive concept. In this paper, the feasibility of transmutation of the long-lived actinide radioactive waste Np-237 with a fusion reactor and a hybrid reactor has been investigated. A new waste management concept of burning HLW (High Level Waste), utilizing released energy and converting Np-237 into fissile fuel Pu-239 through transmutation has been adopted. The detailed neutronics and depletion calculation of waste inventories was carried out with a modified version of one-dimensional neutron transport and burnup calculation code system BISON1.5 in this study. The transmutation rate of Np with relationship to neutron wall loading, Pu and Np with relationship to neutron wall load, Pu and Np concentration in the transmutation zone have been explored as well as relevant results are also given

Neutron personnel dosimetry considerations for fusion reactors

International Nuclear Information System (INIS)

Barton, T.P.; Easterly, C.E.

1979-07-01

The increasing development of fusion reactor technology warrants an evaluation of personnel neutron dosimetry systems to aid in the concurrent development of a radiation protection program. For this reason, current state of knowledge neutron dosimeters have been reviewed with emphasis placed on practical utilization and the problems inherent in each type of dosimetry system. Evaluations of salient parameters such as energy response, latent image instability, and minimum detectable dose equivalent are presented for nuclear emulsion films, track etch techniques, albedo and other thermoluminescent dosimetry techniques, electrical conductivity damage effects, lyoluminescence, thermocurrent, and thermally stimulated exoelectron emission. Brief summaries of dosimetry regulatory requirements and intercomparison study results help to establish compliance and recent trends, respectively. Spectrum modeling data generated by the Neutron Physics Division of Oak Ridge National Laboratory for the Princeton Tokamak Fusion Test Reactor (TFTR) Facility have been analyzed by both International Commission on Radiological Protection fluence to dose conversion factors and an adjoint technique of radiation dosimetry, in an attempt to determine the applicability of current neutron dosimetry systems to deuterium and tritium fusion reactor leakage spectra. Based on the modeling data, a wide range of neutron energies will probably be present in the leakage spectra of the TFTR facility, and no appreciable risk of somatic injury to occupationally exposed workers is expected. The relative dose contributions due to high energy and thermal neutrons indicate that neutron dosimetry will probably not be a serious limitation in the development of fusion power

Reactor Safety Analysis

International Nuclear Information System (INIS)

Arien, B.

2000-01-01

The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

Compact approach to fusion power reactors

International Nuclear Information System (INIS)

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

1984-01-01

The potential of the Reversed-Field Pinch (RFP) for development into an efficient, compact, copper-coil fusion reactor has been quantified by comprehensive parametric tradeoff studies. These compact systems promise to be competitive in size, power density, and cost to alternative energy sources. Conceptual engineering designs that largely substantiate these promising results have since been completed. This 1000-MWe(net) design is described along with a detailed rationale and physics/technology assessment for the compact approach to fusion

Development of 'low activation superconducting wire' for an advanced fusion reactor

International Nuclear Information System (INIS)

Hishinuma, Y.; Yamada, S.; Sagara, A.; Kikuchi, A.; Takeuchi, T.; Matsuda, K.; Taniguchi, H.

2011-01-01

In the D-T burning plasma reactor beyond ITER such as DEMO and fusion power plants assuming the steady-state and long time operation, it will be necessary to consider carefully induced radioactivity and neutron irradiation properties on the all components for fusion reactors. The decay time of the induced radioactivity can control the schedule and scenarios of the maintenance and shutdown on the fusion reactor. V 3 Ga and MgB 2 compound have shorter decay time within 1 years and they will be desirable as a candidate material to realize 'low activation and high magnetic field superconducting magnet' for advanced fusion reactor. However, it is well known that J c -B properties of V 3 Ga and MgB 2 wires are lower than that of the Nb-based A15 compound wires, so the J c -B enhancements on the V 3 Ga and MgB 2 wires are required in order to apply for an advanced fusion reactor. We approached and succeeded to developing the new process in order to improve J c properties of V 3 Ga and MgB 2 wires. In this paper, the recent activities for the J c improvements and detailed new process in V 3 Ga and MgB 2 wires are investigated. (author)

Activation and Radiation Damage Behaviour of Russian Structural Materials for Fusion Reactors in the Fission and Fusion Reactors

International Nuclear Information System (INIS)

Blokhin, A.; Demin, N.; Chernov, V.; Leonteva-Smirnova, M.; Potapenko, M.

2006-01-01

Various structural low (reduced) activated materials have been proposed as a candidate for the first walls-blankets of fusion reactors. One of the main problems connected with using these materials - to minimise the production of long-lived radionuclides from nuclear transmutations and to provide with good technological and functional properties. The selection of materials and their metallurgical and fabrication technologies for fusion reactor components is influenced by this factor. Accurate prediction of induced radioactivity is necessary for the development of the fusion reactor materials. Low activated V-Ti-Cr alloys and reduced activated ferritic-martensitic steels are a leading candidate material for fusion first wall and blanket applications. At the present time a range of compositions and an impurity level are still being investigated to better understand the sensitive of various functional and activation properties to small compositional variations and impurity level. For the two types of materials mentioned above (V-Ti-Cr alloys and 9-12 % Cr f/m steels) and manufactured in Russia (Russia technologies) the analysis of induced activity, hydrogen and helium-production as well as the accumulation of such elements as C, N, O, P, S, Zn and Sn as a function of irradiation time was performed. Materials '' were irradiated '' by fission (BN-600, BOR-60) and fusion (Russian DEMO-C Reactor Project) typical neutron spectra with neutron fluency up to 10 22 n/cm 2 and the cooling time up to 1000 years. The calculations of the transmutation of elements and the induced radioactivity were carried out using the FISPACT inventory code, and the different activation cross-section libraries like the ACDAM, FENDL-2/A and the decay data library FENDL-2/D. It was shown that the level of impurities controls a long-term behaviour of induced activity and contact dose rate for materials. From this analysis the concentration limits of impurities were obtained. The generation of gas