What must Demo do?

International Nuclear Information System (INIS)

Waganer, L.M.; Najmabadi, F.; Tillack, M.S.

1995-01-01

The US fusion demonstration plant (Demo) must satisfy certain top level requirements so that energy producers will confidently invest in a commercial fusion version for their next generation power plant. To instill that level of confidence to both the investor and the public, Demo must achieve high standards in safety, low environmental impact, reliability, and economics. This is a most difficult set of goals to meet. The public is demanding ever more strict environmental rules and regulations. The hazards of radioactive and toxic waste and emissions are becoming better understood. The difficulties of establishing and maintaining long-lived repositories are enormous. Neighborhood action groups have an aversion to large power plants in their back yards. Utilities and independent power producers are reluctant to commit to a long-term financial arrangement for a new technology. To achieve these stringent goals, the competition is continuing to improve to meet these challenges. Only the best can adapt and survive. The Demo plant is not expected to achieve all requirements demanded of the commercial power plant, but it must demonstrate values close enough to the commercial machine so that extrapolation to the commercial carries minimal risk in all key areas. Specifically, Demo must demonstrate all the major performance parameters in an integrated system similar to that of the commercial plant. It should be large enough so that all aspects of the Demo can be confidently scaled to that of the commercial plant, including the economics, reliability, availability, and operability

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)

Fusion power plant studies in Europe

International Nuclear Information System (INIS)

Maisonnier, D.

2007-01-01

The European fusion programme is reactor oriented and it is aimed at the successive demonstration of the scientific, the technological and the economic feasibility of fusion power. For a reactor-oriented fusion development programme, it is essential to have a clear idea of the ultimate goal of the programme, namely a series of models of fusion power plants, in order to define the correct strategy and to assess the pertinence of the on-going activities. The European Power Plant Conceptual Study (PPCS) has been a study of conceptual designs for commercial fusion power plants. It focused on five power plant models, named PPCS A, B, AB, C and D, which are illustrative of a wider spectrum of possibilities. They are all based on the tokamak concept and they have approximately the same net electrical power output, 1500 MWe. These span a range from relatively near-term, based on limited technology and plasma physics extrapolations, to an advanced conception. All five PPCS plant models differ substantially from the models that formed the basis of earlier European studies. They also differ from one another, which lead to differences in economic performance and in the details of safety and environmental impacts. The main emphasis of the PPCS was on system integration. Systems analyses were used to produce self-consistent plant parameter sets with approximately optimal economic characteristics for all models. In the PPCS models, the favourable, inherent, features of fusion have been exploited to provide substantial safety and environmental advantages. The broad features of the safety and environmental conclusions of previous studies have been confirmed and demonstrated with increased confidence. The PPCS study highlighted the need for specific design and R and D activities, in addition to those already underway within the European long term R and D programme, as well as the need to clarify the concept of DEMO, the device that will bridge the gap between ITER and the first

Phenomena Identification and Ranking Table for K-DEMO

Energy Technology Data Exchange (ETDEWEB)

Oh, Kye Min; Kang, Myung Suk; Heo, Gyun Young [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Hyoung Chan [National Fusion Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The purpose of this paper is to describe the current status of the Phenomena Identification Ranking Table (PIRT) for the conceptual design of K-DEMO, Korean Fusion DEMO Plant. K-DEMO is to be planned as the first fusion power plant constructed in South Korea. However, several key technologies such as plasma, materials, and cooling still have large uncertainties. There are also no relevant references to facilitate the design process of K-DEMO due to its different size, commercializing purpose, and regulatory framework. It was proposed to define the phenomena of systems, components, and processes in an accident condition. In this paper, PIRT for K-DEMO was described and analysis based on this tool was performed. We have carried out researches related to safety for fusion power plant in collaboration with the academies funded by NFRI during the past 3 years. As part of this research, Integrated Safety Assessment Methodology (ISAM), which was used to develop GEN-IV nuclear systems, was used to determine the technical safety issues and regulatory requirements for K-DEMO. PIRT is one of ISAM tools. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional research. The results through this tool are expected to contribute on detailed design for K-DEMO as guidance for regulatory requirements and safety systems in the future.

Phenomena Identification and Ranking Table for K-DEMO

International Nuclear Information System (INIS)

Oh, Kye Min; Kang, Myung Suk; Heo, Gyun Young; Kim, Hyoung Chan

2013-01-01

The purpose of this paper is to describe the current status of the Phenomena Identification Ranking Table (PIRT) for the conceptual design of K-DEMO, Korean Fusion DEMO Plant. K-DEMO is to be planned as the first fusion power plant constructed in South Korea. However, several key technologies such as plasma, materials, and cooling still have large uncertainties. There are also no relevant references to facilitate the design process of K-DEMO due to its different size, commercializing purpose, and regulatory framework. It was proposed to define the phenomena of systems, components, and processes in an accident condition. In this paper, PIRT for K-DEMO was described and analysis based on this tool was performed. We have carried out researches related to safety for fusion power plant in collaboration with the academies funded by NFRI during the past 3 years. As part of this research, Integrated Safety Assessment Methodology (ISAM), which was used to develop GEN-IV nuclear systems, was used to determine the technical safety issues and regulatory requirements for K-DEMO. PIRT is one of ISAM tools. It can recognize vulnerabilities of systems and identify the gaps in technical areas requiring additional research. The results through this tool are expected to contribute on detailed design for K-DEMO as guidance for regulatory requirements and safety systems in the future

Safety research on fusion DEMO in Japan: Toward development of safety strategy of a water-cooled DEMO

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Makoto, E-mail: nakamura.makoto@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Tobita, Kenji; Someya, Youji; Utoh, Hiroyasu; Sakamoto, Yoshiteru [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Gulden, Werner [Fusion for Energy, Garching D-85748 (Germany)

2016-11-01

Highlights: • This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. • We report analyses of two transients: (i) complete loss of decay heat removal and (ii) major ex-VV LOCA. • The MELCOR analysis has clarified the temperature histories of the DEMO components in complete loss of decay heat removal. • A strategy to reduce the pressure load to the final barrier confining radioactive materials is proposed against the major ex-VV LOCA. - Abstract: This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. A basic strategy of development of the safety guidelines is described for DEMO based on a water-cooled solid pebble bed blanket. Clarification of safety features of the DEMO in accident situations is a key issue to develop the guidelines. Recent achievements in understanding of the safety features of the water-cooled DEMO are reported. The MELCOR analysis has clarified the temperature histories of the DEMO components in a complete loss of decay heat removal event. The transient behavior of the first wall temperature is found to be essentially different from that of ITER. The pressure load to the tokamak cooling water system vault (TCWSV) is analyzed based on a simple model equation of the energy conservation. If the amount of the primary coolant is the same as that of Slim-CS, the previous small Japanese DEMO, the discharged water does not damage the TCWSV with the volume and pressure-tightness similar to those of pressurized light water reactors. It is shown that implementation of a pressure suppression system to the small TCWSV is effective to suppress the pressure load to the second confinement barrier.

Development of DEMO-FNS tokamak for fusion and hybrid technologies

Science.gov (United States)

Kuteev, B. V.; Azizov, E. A.; Alexeev, P. N.; Ignatiev, V. V.; Subbotin, S. A.; Tsibulskiy, V. F.

2015-07-01

The history of fusion-fission hybrid systems based on a tokamak device as an extremely efficient DT-fusion neutron source has passed through several periods of ample research activity in the world since the very beginning of fusion research in the 1950s. Recently, a new roadmap of the hybrid program has been proposed with the goal to build a pilot hybrid plant (PHP) in Russia by 2030. Development of the DEMO-FNS tokamak for fusion and hybrid technologies, which is planned to be built by 2023, is the key milestone on the path to the PHP. This facility is in the phase of conceptual design aimed at providing feasibility studies for a full set of steady state tokamak technologies at a fusion energy gain factor Q ˜ 1, fusion power of ˜40 MW and opportunities for testing a wide range of hybrid technologies with the emphasis on continuous nuclide processing in molten salts. This paper describes the project motivations, its current status and the key issues of the design.

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.)

A preliminary conceptual design study for Korean fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Kim, Hyoung Chan; Oh, Sangjun; Lee, Young Seok; Yeom, Jun Ho; Im, Kihak; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Kessel, Charles; Brown, Thomas; Titus, Peter [Princeton Plasma Physics Laboratory, Princeton University, Princeton, NJ 08543 (United States)

2013-10-15

Highlights: ► Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. ► Present design guidelines and requirements of Korean DEMO reactor. ► Present a preliminary design of TF (toroidal field) and CS (central solenoid) magnet. ► Present a preliminary result of the radial build scheme of Korean DEMO reactor. -- Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, considering practical engineering feasibilities. By using high performance Nb{sub 3}Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters.

Technology and Plasma Physics Developments needed for DEMO

International Nuclear Information System (INIS)

Lackner, K.

2006-01-01

Although no universally agreed definition of the next step after ITER exists at present it is commonly accepted that significant progress beyond the ITER base-line operating physics modes and the technologies employed in it are needed. We first review the role of DEMO in the different proposed fusion road maps and derive from them the corresponding performance requirements. A fast track to commercial fusion implies that DEMO is already close to a first of a kind power plant in all aspects except average availability. Existing power plant studies give therefore also a good approximation to the needs of DEMO. We outline the options for achieving the needed physics progress in the different characteristic parameters, and the implications for the experimental programme of ITER and accompanying satellite devices. On the time scale of the operation of ITER and of the planning DEMO, ab-initio modelling of fusion plasmas is also expected to assume a qualitatively new role. Besides the mapping of the reactor regime of plasma physics and the integration of a burning plasma with the principal reactor technologies on ITER, the development of functional and structural materials capable of handling the high power fluxes and neutron fluences, respectively is also on the critical path to DEMO. Finally we discuss the potential contributions of other confinement concepts (stellarators and spherical tokamaks) to the design of DEMO. (author)

Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

2015-10-15

Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

Comparative evaluation of remote maintenance schemes for fusion DEMO reactor

International Nuclear Information System (INIS)

Utoh, Hiroyasu; Tobita, Kenji; Someya, Youji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto

2015-01-01

Highlights: • Various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. • The banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme. • The key engineering issues are in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability. - Abstract: Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

Development of Tokamak Reactor System Code and Performance for Early Realization of DEMO

International Nuclear Information System (INIS)

Hong, B. G.; Lee, D. W.; Kim, Y.

2006-01-01

To develop the concepts of DEMO and identify the design parameters, dependence on performance objectives, design features and physical and technical constraints have to be considered. System analyses are necessary to find device variables which optimize figures of merit such as major radius, ignition margin, divertor heat load, neutron wall load, etc. Demonstration fusion power plant, DEMO is regarded as the last step before the development of a commercial fusion reactor in Korea National Basic Plan for the Development of Fusion Energy. The DEMO should demonstrate a net electric power generation, a tritium self sufficiency, and the safety aspect of a power plant. Performance of DEMO for early realization has been investigated with a limited extension from the plasma physics and technology in the 2nd phase of the ITER operation (EPP phase)

Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2016-05-15

To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

Demonstration tokamak fusion power plant for early realization of net electric power generation

International Nuclear Information System (INIS)

Hiwatari, R.; Okano, K.; Asaoka, Y.; Shinya, K.; Ogawa, Y.

2005-01-01

A demonstration tokamak fusion power plant Demo-CREST is proposed as the device for early realization of net electric power generation by fusion energy. The plasma configuration for Demo-CREST is optimized to satisfy the electric breakeven condition (the condition for net electric power, P e net = 0 MW) with the plasma performance of the ITER reference operation mode. This optimization method is considered to be suitable for the design of a demonstration power plant for early realization of net electric power generation, because the demonstration power plant has to ensure the net electric generation. Plasma performance should also be more reliably achieved than in past design studies. For the plasma performance planned in the present ITER programme, net electric power from 0 to 500 MW is possible with Demo-CREST under the following engineering conditions: maximum magnetic field 16 T, thermal efficiency 30%, NBI system efficiency 50% and NBI current drive power restricted to 200 MW. By replacing the blanket system with one of higher thermal efficiency, a net electric power of about 1000 MW is also possible so that the performance of the commercial plant with Demo-CREST can also be studied from the economic point of view. The development path from the experimental reactor 'ITER' to the commercial plant 'CREST' through the demonstration power plant 'Demo-CREST' is proposed as an example of the fast track concept. (author)

Diagnostics for plasma control on DEMO: challenges of implementation

NARCIS (Netherlands)

Donne, A. J. H.; Costley, A. E.; Morris, A. W.

2012-01-01

As a test fusion power plant, DEMO will have to demonstrate reliability and very long pulse/steady-state operation, which calls for unprecedented robustness and reliability of all diagnostic systems (also requiring adequate redundancy). But DEMO will have higher levels of neutron and gamma fluxes,

Research and development plan of fusion technologies in JAERI toward DEMO reactors

International Nuclear Information System (INIS)

Nishitani, Takeo; Hayashi, Takumi; Abe, Tetsuya; Akiba, Masato; Isono, Takaaki; Inoue, Takashi; Enoeda, Mikio; Okuno, Kiyoshi; Koizumi, Norikiyo; Sakamoto, Keishi; Sato, Satoshi; Jitsukawa, Shiro; Sugimoto, Masayoshi; Suzuki, Satoshi; Seki, Shogo; Takatsu, Hideyuki; Tanzawa, Sadamitsu; Tsuchiya, Kunihiko; Nishi, Masataka; Hayashi, Kimio; Matsui, Hideki; Yamanishi, Toshihiko; Watanabe, Kazuhiro

2005-03-01

In accordance with the 'Third Phase Basic Program on Fusion Research and Development' established by the Fusion Council of the Japan Atomic Energy Commission, research and development (R and D) of fusion technologies aim at realization of two elements: development of ITER key components and their improvement for higher performances; and construction of sound technical basis of fusion nuclear technologies essential for fusion energy utilization. JAERI has been assigned in the Third Phase Basic Program as a responsible institute for developing the above two elements, and accordingly has been implementing technology R and Ds categorized in the following three areas: R and D for ITER construction and operation; R and D for ITER utilization (blanket testing in ITER) and toward DEMO; and R and D on basic fusion technologies. The present report reviews the status and the plan of fusion technology R and Ds in the latter two areas, and presents the technical objectives, technical issues, status of R and D and near-term R and D plans for: breeding blankets; structural materials; the IFMIF program; improvements of the key ITER components for higher performances toward DEMO; and basic fusion technologies. (author)

Objective Provision Tree for K-DEMO

International Nuclear Information System (INIS)

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

2014-01-01

In current nuclear field based on technology-neutral approach, safety principles and design have been considered for Generation IV (Gen-IV) nuclear power plants in parallel. This strategy can save resource, time, and manpower while keeping achievable safety. For this reason, the studies related with safety affecting significant design parameters for planned construction or fusion plants was needed and required even though K-DEMO is staying in pre-conceptual design phase. Objective Provision Tree (OPT) is one of the tools of Integrated Safety Assessment Methodology (ISAM) developed by Risk and Safety Working Group (RSWG) for design and assessment of Gen-IV. This is suitable tool to recognize and investigate safety issues from previous engineering experience. The purpose of this paper is to suggest multiple barriers/critical safety function and to describe the current status of the OPT for the conceptual design of K-DEMO. In this paper, critical safety functions were defined and OPT for K-DEMO was described and performed. We have carried out researches related to safety for fusion power plant in collaboration with the academies funded by NFRI during the past 4 years. As part of this research, Integrated Safety Assessment Methodology (ISAM), which was used to develop GEN-IV nuclear systems, was used to determine the technical safety issues and regulatory requirements for K-DEMO. OPT is one of ISAM tools

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-10-15

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

Preliminary concept design of the divertor remote handling system for DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2014-11-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

Conceptual design of the beam source for the DEMO Neutral Beam Injectors

Science.gov (United States)

Sonato, P.; Agostinetti, P.; Fantz, U.; Franke, T.; Furno, I.; Simonin, A.; Tran, M. Q.

2016-12-01

DEMO (DEMOnstration Fusion Power Plant) is a proposed nuclear fusion power plant that is intended to follow the ITER experimental reactor. The main goal of DEMO will be to demonstrate the possibility to produce electric energy from the fusion reaction. The injection of high energy neutral beams is one of the main tools to heat the plasma up to fusion conditions. A conceptual design of the Neutral Beam Injector (NBI) for the DEMO fusion reactor, is currently being developed by Consorzio RFX in collaboration with other European research institutes. High efficiency and low recirculating power, which are fundamental requirements for the success of DEMO, have been taken into special consideration for the DEMO NBI. Moreover, particular attention has been paid to the issues related to reliability, availability, maintainability and inspectability. A conceptual design of the beam source for the DEMO NBI is here presented featuring 20 sub-sources (two adjacent columns of 10 sub-sources each), following a modular design concept, with each sub-source featuring its radio frequency driver, capable of increasing the reliability and availability of the DEMO NBI. Copper grids with increasing size of the apertures have been adopted in the accelerator, with three main layouts of the apertures (circular apertures, slotted apertures and frame-like apertures for each sub-source). This design, permitting to significantly decrease the stripping losses in the accelerator without spoiling the beam optics, has been investigated with a self-consistent model able to study at the same time the magnetic field, the electrostatic field and the trajectory of the negative ions. Moreover, the status on the R&D carried out in Europe on the ion sources is presented.

Overview of the European Union fusion nuclear technologies development and essential elements on the way to DEMO

International Nuclear Information System (INIS)

Andreani, R.; Diegele, E.; Gulden, W.; Laesser, R.; Maisonnier, D.; Murdoch, D.; Pick, M.; Poitevin, Y.

2006-01-01

EU is strongly preparing ITER construction involving the system of EU Associations, universities and industry. The European programme has been steered to be in line with the present conception of a future power reactor. Thirty percent of the fusion research budget has been spent on long-term related activities managed by EFDA. These include Power Plant Conceptual Studies (PPCS), the recently undertaken DEMO Conceptual Studies, design and R and D for breeder blankets, low activation materials and IFMIF. Developments on fuel cycle, neutronics, safety and socio-economics complement those specifically performed for ITER. Two EU helium-cooled DEMO blankets will be tested in ITER, using liquid lithium-lead and solid ceramics as breeders. The blanket structures will use EUROFER. Irradiations to 70-80 dpa will qualify EUROFER for DEMO. Advanced materials, in particular SiC f /SiC, under development, could provide more thermodynamically efficient blankets. Even with a fully successful ITER, a number of issues will remain open in technology. The application of high temperature superconductors, essential progress in materials, blanket design and remote handling, are required to produce environmentally safe and economically competitive fusion. A fully integrated world wide international programme is the best way to efficiently progress in these fields

Neutronic analyses and tools development efforts in the European DEMO programme

Energy Technology Data Exchange (ETDEWEB)

Fischer, U., E-mail: ulrich.fischer@kit.edu [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Bachmann, C. [European Fusion Development Agreement (EFDA), Garching (Germany); Bienkowska, B. [Association IPPLM-Euratom, IPPLM Warsaw/INP Krakow (Poland); Catalan, J.P. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Drozdowicz, K.; Dworak, D. [Association IPPLM-Euratom, IPPLM Warsaw/INP Krakow (Poland); Leichtle, D. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Fusion for Energy (F4E), Barcelona (Spain); Lengar, I. [MESCS-JSI, Ljubljana (Slovenia); Jaboulay, J.-C. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Lu, L. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Moro, F. [Associazione ENEA-Euratom, ENEA Fusion Division, Frascati (Italy); Mota, F. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Sanz, J. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Szieberth, M. [Budapest University of Technology and Economics (BME), Budapest (Hungary); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pampin, R. [Fusion for Energy (F4E), Barcelona (Spain); Porton, M. [Euratom/CCFE Fusion Association, Culham Science Centre for Fusion Energy (CCFE), Culham (United Kingdom); Pereslavtsev, P. [Association KIT-Euratom, Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Ogando, F. [Universidad Nacional de Educación a Distancia (UNED), Madrid (Spain); Rovni, I. [Budapest University of Technology and Economics (BME), Budapest (Hungary); and others

2014-10-15

Highlights: •Evaluation of neutronic tools for application to DEMO nuclear analyses. •Generation of a DEMO model for nuclear analyses based on MC calculations. •Nuclear analyses of the DEMO reactor equipped with a HCLL-type blanket. -- Abstract: The European Fusion Development Agreement (EFDA) recently launched a programme on Power Plant Physics and Technology (PPPT) with the aim to develop a conceptual design of a fusion demonstration reactor (DEMO) addressing key technology and physics issues. A dedicated part of the PPPT programme is devoted to the neutronics which, among others, has to define and verify requirements and boundary conditions for the DEMO systems. The quality of the provided data depends on the capabilities and the reliability of the computational tools. Accordingly, the PPPT activities in the area of neutronics include both DEMO nuclear analyses and development efforts on neutronic tools including their verification and validation. This paper reports on first neutronics studies performed for DEMO, and on the evaluation and further development of neutronic tools.

Neutronic analyses and tools development efforts in the European DEMO programme

International Nuclear Information System (INIS)

Fischer, U.; Bachmann, C.; Bienkowska, B.; Catalan, J.P.; Drozdowicz, K.; Dworak, D.; Leichtle, D.; Lengar, I.; Jaboulay, J.-C.; Lu, L.; Moro, F.; Mota, F.; Sanz, J.; Szieberth, M.; Palermo, I.; Pampin, R.; Porton, M.; Pereslavtsev, P.; Ogando, F.; Rovni, I.

2014-01-01

Highlights: •Evaluation of neutronic tools for application to DEMO nuclear analyses. •Generation of a DEMO model for nuclear analyses based on MC calculations. •Nuclear analyses of the DEMO reactor equipped with a HCLL-type blanket. -- Abstract: The European Fusion Development Agreement (EFDA) recently launched a programme on Power Plant Physics and Technology (PPPT) with the aim to develop a conceptual design of a fusion demonstration reactor (DEMO) addressing key technology and physics issues. A dedicated part of the PPPT programme is devoted to the neutronics which, among others, has to define and verify requirements and boundary conditions for the DEMO systems. The quality of the provided data depends on the capabilities and the reliability of the computational tools. Accordingly, the PPPT activities in the area of neutronics include both DEMO nuclear analyses and development efforts on neutronic tools including their verification and validation. This paper reports on first neutronics studies performed for DEMO, and on the evaluation and further development of neutronic tools

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

Directory of Open Access Journals (Sweden)

Laila El-Guebaly

2016-08-01

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

Conceptual study of ECH/ECCD system for fusion DEMO plant

International Nuclear Information System (INIS)

Sakamoto, K.; Takahashi, K.; Kasugai, A.; Minami, R.; Kobayashi, N.; Nishio, S.; Sato, M.; Tobita, K.

2006-01-01

The conceptual study of the electron cyclotron heating and current drive (ECH/ECCD) system for a DEMO reactor was carried out. The ECH/ECCD system was considered on the basis of a design of the DEMO reactor by JAERI. The reactor is a low aspect ratio tokamak, and its size and magnetic field are similar to those of ITER. Therefore, many ECH/ECCD technologies developed at 170 GHz for ITER can be applied. Truly continuous operation is needed for DEMO, and the neutron fluence from the plasma is two orders of magnitude higher than that of ITER. An RF launcher that has reliability under the condition of high neutron fluence is, critically, important. For power deposition control in the plasma, a gyrotron frequency tuning system is considered as the primary candidate to realize a simple and robust launching system, but two RF beam steering systems are discussed as alternatives

First generation of fusion power plants: Design and technology. Proceedings of the 2. IAEA technical meeting

International Nuclear Information System (INIS)

2008-01-01

This series of meetings has been initiated under recommendation of the International Fusion Research Council for the IAEA and is expected to initiate, develop and mature ideas on fusion strategy that would be of benefit for all players. The present objectives of this meeting are to provide a forum to discuss concepts, technology and environmental aspects of future fusion power plants, the next step following ITER, their role in future energy mix and to assess a selection of urgent topics aiming at identifying the physics and the technological requirements that ITER and a fusion grade materials developing programme will have to address to support the construction of a DEMO(s) fusion power plant(s) prototype demonstrating viable economics. The meeting was organized in five sessions addressing five topics: - (PPCA) Power Plant Concepts and systems Analysis. - (MCP) Materials analysis/Components design/Plasma requirements - (NE) Non-Electric applications of fusion - (SESE) Social, Economic, Safety and Environmental aspects of fusion - (EP) Energy Policy, strategy and scenario for fusion development. A summary session took place at the end of the meeting. Thirty-three participants representing 12 Countries and 3 International Organizations were present at the meeting

DEMO diagnostics and burn control

NARCIS (Netherlands)

Biel, W.; De Baar, M.; Dinklage, A.; Felici, F.; König, R.; Meister, H.; Treutterer, W.; Wenninger, R.

2015-01-01

The development of the control system for a tokamak demonstration fusion reactor (DEMO) faces unprecedented challenges. First, the requirements for control reliability and accuracy are more stringent than on existing fusion devices: any loss of plasma control on DEMO may result in a disruption which

Neutronic performance optimization study of Indian fusion demo reactor first wall and breeding blanket

International Nuclear Information System (INIS)

Swami, H.L.; Danani, C.

2015-01-01

In frame of design studies of Indian Nuclear Fusion DEMO Reactor, neutronic performance optimization of first wall and breeding blanket are carried out. The study mainly focuses on tritium breeding ratio (TBR) and power density responses estimation of breeding blanket. Apart from neutronic efficiency of existing breeding blanket concepts for Indian DEMO i.e. lead lithium ceramic breeder and helium cooled solid breeder concept other concepts like helium cooled lead lithium and helium-cooled Li_8PbO_6 with reflector are also explored. The aim of study is to establish a neutronically efficient breeding blanket concept for DEMO. Effect of first wall materials and thickness on breeding blanket neutronic performance is also evaluated. For this study 1 D cylindrical neutronic model of DEMO has been constructed according to the preliminary radial build up of Indian DEMO. The assessment is being done using Monte Carlo based radiation transport code and nuclear cross section data file ENDF/B- VII. (author)

Conceptual design of the DEMO neutral beam injectors: main developments and R&D achievements

Science.gov (United States)

Sonato, P.; Agostinetti, P.; Bolzonella, T.; Cismondi, F.; Fantz, U.; Fassina, A.; Franke, T.; Furno, I.; Hopf, C.; Jenkins, I.; Sartori, E.; Tran, M. Q.; Varje, J.; Vincenzi, P.; Zanotto, L.

2017-05-01

The objectives of the nuclear fusion power plant DEMO, to be built after the ITER experimental reactor, are usually understood to lie somewhere between those of ITER and a ‘first of a kind’ commercial plant. Hence, in DEMO the issues related to efficiency and RAMI (reliability, availability, maintainability and inspectability) are among the most important drivers for the design, as the cost of the electricity produced by this power plant will strongly depend on these aspects. In the framework of the EUROfusion Work Package Heating and Current Drive within the Power Plant Physics and Development activities, a conceptual design of the neutral beam injector (NBI) for the DEMO fusion reactor has been developed by Consorzio RFX in collaboration with other European research institutes. In order to improve efficiency and RAMI aspects, several innovative solutions have been introduced in comparison to the ITER NBI, mainly regarding the beam source, neutralizer and vacuum pumping systems.

Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

Science.gov (United States)

Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

2016-01-01

Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

Pre-conceptual studies and R and D for DEMO superconducting magnets

Energy Technology Data Exchange (ETDEWEB)

Bruzzone, Pierluigi, E-mail: pierluigi.bruzzone@psi.ch

2014-10-15

Highlights: • Comparison of DEMO parameters vs. ITER for TF coils. • Hybridization of winding pack, Nb/Nb{sub 3}Sn, by graded layer winding. • Use of react and wind method opposite to wind and react with related advantages. • Feasibility, reliability and cost competitiveness for DEMO. - Abstract: The DEMO plant will demonstrate by mid century the feasibility of electric power generation by nuclear fusion. Since 2011, conceptual design studies are coordinated by the EFDA Power Plant Physics and Technology (PPPT) Division, with the aim of identifying requirements, propose design approaches and start R and D for the magnet system of DEMO. The input and generic boundary conditions are given by the system codes: the major radius of the tokamak is about 9 m. The proposed operating current at 13.6 T peak field is 82 kA, placing the DEMO TF conductor at substantially higher performance compared to ITER TF (68 kA/11.5 T). The innovative winding layout is a graded, layer wound with Nb{sub 3}Sn/NbTi hybridization, aiming at minimizing the size and the cost of the superconductor. Two options are considered for the Nb{sub 3}Sn conductor: one a “wind and react” cable-in-conduit (CICC) with reduced void fraction and rectangular shape. The other conductor is a “react and wind” flat cable with copper segregation and thick steel conduit assembled by longitudinal weld. The conductor designs were first drafted in 2012 and updated in 2013 based on a first round of assessments, which includes electromagnetic, thermal-hydraulic and mechanical analysis. The manufacture of full size prototype conductors is planned in 2014. The technical requirement of the DEMO superconducting magnets is highlighted in comparison to ITER and other fusion devices. The large size of the DEMO tokamak is the main challenge for the demonstration of the feasibility of power generation by fusion. Together with the technical issues, the cost of the superconducting magnets will be eventually the

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)

Design concept of K-DEMO for near-term implementation

Science.gov (United States)

Kim, K.; Im, K.; Kim, H. C.; Oh, S.; Park, J. S.; Kwon, S.; Lee, Y. S.; Yeom, J. H.; Lee, C.; Lee, G.-S.; Neilson, G.; Kessel, C.; Brown, T.; Titus, P.; Mikkelsen, D.; Zhai, Y.

2015-05-01

A Korean fusion energy development promotion law (FEDPL) was enacted in 2007. As a following step, a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. After the thorough 0D system analysis, the parameters of the main machine characterized by the major and minor radii of 6.8 and 2.1 m, respectively, were chosen for further study. The analyses of heating and current drives were performed for the development of the plasma operation scenarios. Preliminary results on lower hybrid and neutral beam current drive are included herein. A high performance Nb3Sn-based superconducting conductor is adopted, providing a peak magnetic field approaching 16 T with the magnetic field at the plasma centre above 7 T. Pressurized water is the prominent choice for the main coolant of K-DEMO when the balance of plant development details is considered. The blanket system adopts a ceramic pebble type breeder. Considering plasma performance, a double-null divertor is the reference configuration choice of K-DEMO. For a high availability operation, K-DEMO incorporates a design with vertical maintenance. A design concept for K-DEMO is presented together with the preliminary design parameters.

Sensitivity and uncertainty analysis for the tritium breeding ratio of a DEMO fusion reactor with a helium cooled pebble bed blanket

Directory of Open Access Journals (Sweden)

Nunnenmann Elena

2017-01-01

Full Text Available An uncertainty analysis was performed for the tritium breeding ratio (TBR of a fusion power plant of the European DEMO type using the MCSEN patch to the MCNP Monte Carlo code. The breeding blanket was of the type Helium Cooled Pebble Bed (HCPB, currently under development in the European Power Plant Physics and Technology (PPPT programme for a fusion power demonstration reactor (DEMO. A suitable 3D model of the DEMO reactor with HCPB blanket modules, as routinely used for blanket design calculations, was employed. The nuclear cross-section data were taken from the JEFF-3.2 data library. For the uncertainty analysis, the isotopes H-1, Li-6, Li-7, Be-9, O-16, Si-28, Si-29, Si-30, Cr-52, Fe-54, Fe-56, Ni-58, W-182, W-183, W-184 and W-186 were considered. The covariance data were taken from JEFF-3.2 where available. Otherwise a combination of FENDL-2.1 for Li-7, EFF-3 for Be-9 and JENDL-3.2 for O-16 were compared with data from TENDL-2014. Another comparison was performed with covariance data from JEFF-3.3T1. The analyses show an overall uncertainty of ± 3.2% for the TBR when using JEFF-3.2 covariance data with the mentioned additions. When using TENDL-2014 covariance data as replacement, the uncertainty increases to ± 8.6%. For JEFF-3.3T1 the uncertainty result is ± 5.6%. The uncertainty is dominated by O-16, Li-6 and Li-7 cross-sections.

Reliability and availability requirements analysis for DEMO: fuel cycle system

International Nuclear Information System (INIS)

Pinna, T.; Borgognoni, F.

2015-01-01

The Demonstration Power Plant (DEMO) will be a fusion reactor prototype designed to demonstrate the capability to produce electrical power in a commercially acceptable way. Two of the key elements of the engineering development of the DEMO reactor are the definitions of reliability and availability requirements (or targets). The availability target for a hypothesized Fuel Cycle has been analysed as a test case. The analysis has been done on the basis of the experience gained in operating existing tokamak fusion reactors and developing the ITER design. Plant Breakdown Structure (PBS) and Functional Breakdown Structure (FBS) related to the DEMO Fuel Cycle and correlations between PBS and FBS have been identified. At first, a set of availability targets has been allocated to the various systems on the basis of their operating, protection and safety functions. 75% and 85% of availability has been allocated to the operating functions of fuelling system and tritium plant respectively. 99% of availability has been allocated to the overall systems in executing their safety functions. The chances of the systems to achieve the allocated targets have then been investigated through a Failure Mode and Effect Analysis and Reliability Block Diagram analysis. The following results have been obtained: 1) the target of 75% for the operations of the fuelling system looks reasonable, while the target of 85% for the operations of the whole tritium plant should be reduced to 80%, even though all the tritium plant systems can individually reach quite high availability targets, over 90% - 95%; 2) all the DEMO Fuel Cycle systems can reach the target of 99% in accomplishing their safety functions. (authors)

Considerations on the DEMO pellet fuelling system

Energy Technology Data Exchange (ETDEWEB)

Lang, P.T., E-mail: peter.lang@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Day, Ch. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Fable, E. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Igitkhanov, Y. [Karlsruhe Institute of Technology, 76021 Karlsruhe (Germany); Köchl, F. [Association EURATOM-Ö AW/ATI, Atominstitut, TU Wien, 1020 Vienna (Austria); Mooney, R. [Culham Centre for Fusion Energy, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Pegourie, B. [CEA, IRFM, 13108 Saint-Paul-lez-Durance (France); Ploeckl, B. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); Wenninger, R. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany); EFDA, Garching (Germany); Zohm, H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

2015-10-15

Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

Considerations on the DEMO pellet fuelling system

International Nuclear Information System (INIS)

Lang, P.T.; Day, Ch.; Fable, E.; Igitkhanov, Y.; Köchl, F.; Mooney, R.; Pegourie, B.; Ploeckl, B.; Wenninger, R.; Zohm, H.

2015-01-01

Graphical abstract: - Highlights: • Considerations are made for a core particle fuelling system covering all DEMO requirements. • Particle deposition beyond the pedestal top is needed to achieve efficient fuelling. • Conventional pellet technology enabling launching from the torus inboard side can be used. • Efforts have been taken for integrating a suitable pellet guiding system into the EU DEMO model. • In addition, further techniques bearing potential for advanced fuelling performance are considered. - Abstract: The Demonstration Fusion Power Reactor DEMO is the step foreseen to bridge the gap between ITER and the first commercial fusion power plant. One key element in the European work plan for DEMO is the elaboration of a conceptual design for a suitable core particle fuelling system. First considerations for such a system are presented in this contribution. Following the well-considered ITER solution, most analysis performed in this study assumes conventional pellet technology will be used for the fuelling system. However, taking advantage of the less compressed time frame for the DEMO project, several other techniques thought to bear potential for advanced fuelling performance are considered as well. In a first, basic analysis all actuation parameters at hand and their implications on the fuelling performance were considered. Tentative transport modeling of a reference scenario strongly indicates only particles deposited inside the plasma pedestal allow for efficient fuelling. Shallow edge fuelling results in an unbearable burden on the fuel cycle. Sufficiently deep particle deposition seems technically achievable, provided pellets are launched from the torus inboard at sufficient speed. All components required for a DEMO pellet system capable for high speed inboard pellet launch are already available or can be developed in due time with reasonable efforts. Furthermore, steps to integrate this solution into the EU DEMO model are taken.

RAMI analysis for DEMO HCPB blanket concept cooling system

Energy Technology Data Exchange (ETDEWEB)

Dongiovanni, Danilo N., E-mail: danilo.dongiovanni@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Pinna, Tonio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati (Italy); Carloni, Dario [KIT, Institute of Neutron Physics and Reactor Technology (INR) – KIT (Germany)

2015-10-15

Highlights: • RAMI (reliability, availability, maintainability and inspectability) preliminary assessment for HCPB blanket concept cooling system. • Reliability block diagram (RBD) modeling and analysis for HCPB primary heat transfer system (PHTS), coolant purification system (CPS), pressure control system (PCS), and secondary cooling system. • Sensitivity analysis on system availability performance. • Failure models and repair models estimated on the base of data from the ENEA fusion component failure rate database (FCFRDB). - Abstract: A preliminary RAMI (reliability, availability, maintainability and inspectability) assessment for the HCPB (helium cooled pebble bed) blanket cooling system based on currently available design for DEMO fusion power plant is presented. The following sub-systems were considered in the analysis: blanket modules, primary cooling loop including pipework and steam generators lines, pressure control system (PCS), coolant purification system (CPS) and secondary cooling system. For PCS and CPS systems an extrapolation from ITER Test Blanket Module corresponding systems was used as reference design in the analysis. Helium cooled pebble bed (HCPB) system reliability block diagrams (RBD) models were implemented taking into account: system reliability-wise configuration, operating schedule currently foreseen for DEMO, maintenance schedule and plant evolution schedule as well as failure and corrective maintenance models. A simulation of plant activity was then performed on implemented RBDs to estimate plant availability performance on a mission time of 30 calendar years. The resulting availability performance was finally compared to availability goals previously proposed for DEMO plant by a panel of experts. The study suggests that inherent availability goals proposed for DEMO PHTS system and Tokamak auxiliaries are potentially achievable for the primary loop of the HCPB concept cooling system, but not for the secondary loop. A

DEMO port plug design and integration studies

Science.gov (United States)

Grossetti, G.; Boccaccini, L. V.; Cismondi, F.; Del Nevo, A.; Fischer, U.; Franke, T.; Granucci, G.; Hernández, F.; Mozzillo, R.; Strauß, D.; Tran, M. Q.; Vaccaro, A.; Villari, R.

2017-11-01

The EUROfusion Consortium established in 2014 and composed by European Fusion Laboratories, and in particular the Power Plant Physics and Technology department aims to develop a conceptual design for the Fusion DEMOnstration Power Plant, DEMO. With respect to present experimental machines and ITER, the main goals of DEMO are to produce electricity continuously for a period of about 2 h, with a net electrical power output of a few hundreds of MW, and to allow tritium self-sufficient breeding with an adequately high margin in order to guarantee its planned operational schedule, including all planned maintenance intervals. This will eliminate the need to import tritium fuel from external sources during operations. In order to achieve these goals, extensive engineering efforts as well as physics studies are required to develop a design that can ensure a high level of plant reliability and availability. In particular, interfaces between systems must be addressed at a very early phase of the project, in order to proceed consistently. In this paper we present a preliminary design and integration study, based on physics assessments for the EU DEMO1 Baseline 2015 with an aspect ratio of 3.1 and 18 toroidal field coils, for the DEMO port plugs. These aim to host systems like electron cyclotron heating launchers currently developed within the Work Package Heating and Current Drive that need an external radial access to the plasma and through in-vessel systems like the breeder blanket. A similar approach shown here could be in principle followed by other systems, e.g. other heating and current drive systems or diagnostics. The work addresses the interfaces between the port plug and the blanket considering the helium-cooled pebble bed and the water cooled lithium lead which are two of four breeding blanket concepts under investigation in Europe within the Power Plant Physics and Technology Programme: the required openings will be evaluated in terms of their impact onto the

Development step toward fusion power plant and role of experimental reactor ITER

International Nuclear Information System (INIS)

Hiwatari, Ryouji; Asaoka, Yoshiyuki; Okano, Kunihiko

2005-01-01

The development of fusion energy is going into the experimental reactor stage, and the thermal energy from the fusion reaction will be generated in a plant scale through the ITER (International Thermonuclear Experimental Reactor) project. The remaining critical issue toward the realization of fusion energy is to map out the development strategy. Recently early realization approach as for the fusion energy development is being discussed in Japan, Europe, and the United States. This approach implies that the devices for a Demo reactor and a proto-type reactor as seen in the fast breeder reactor are combined into a single device in order to advance the fusion energy development. On the other hand, a clear development road map for fusion energy hasn't been suggested yet, and whether that early realization approach is feasible or not is still ambiguous. In order to realize the fusion energy as an user-friendly energy system, the suggestion of the development missions and the road map from the user-side point of view is instructive not only to Japanese but also to other country's development policy after the ITER project. In this report, first of all, the development missions from the user's point of view have been structured. Second, the development target required to demonstrate net electric generation and to introduce the fusion energy into the market is investigated, respectively. This investigation reveals that the completion of the ITER reference operation gives the outlook toward the demonstration of net electric generation and that the completion of the ITER advanced operation gives the possibility to introduce the fusion energy into the market. At last, the electric demonstration power plant Demo-CREST and the commercial power plant CREST are proposed to construct the development road map for fusion energy. (author)

Global shutdown dose rate maps for a DEMO conceptual design

International Nuclear Information System (INIS)

Leichtle, D.; Pereslavtsev, P.; Sanz, J.; Catalan, J.P.; Juarez, R.

2015-01-01

Highlights: • Application of R2S-method on high-resolution full torus sector mesh for DEMO. • Absorbed dose rates after shutdown for a variely of RH equipment at typical locations. • Idenification of radiation levels at several port based locations. - Abstract: For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU. The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1 GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

Global shutdown dose rate maps for a DEMO conceptual design

Energy Technology Data Exchange (ETDEWEB)

Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pereslavtsev, P. [Karlsruhe Institute of Technology KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Sanz, J.; Catalan, J.P.; Juarez, R. [Universidad Nacional de Educación a Distancia(UNED), E.T.S. Ingenieros Industriales, C/ Juan del Rosal 12, 28040 Madrid (Spain)

2015-10-15

Highlights: • Application of R2S-method on high-resolution full torus sector mesh for DEMO. • Absorbed dose rates after shutdown for a variely of RH equipment at typical locations. • Idenification of radiation levels at several port based locations. - Abstract: For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU. The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1 GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

Japanese endeavors to establish technological bases for DEMO

Energy Technology Data Exchange (ETDEWEB)

Yamada, Hiroshi, E-mail: yamada.hiroshi@nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Kasada, Ryuta [Institute of Advanced Energy, Kyoto University, Uji, Kyoto 611-0011 (Japan); Ozaki, Akira [Japan Atomic Industrial Forum, Inc., Minato-ku, Tokyo 105-8605 (Japan); Sakamoto, Ryuichi [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Sakamoto, Yoshiteru [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Takenaga, Hidenobu [Naka Fusion Institute, Japan Atomic Energy Agency, Naka, Ibaraki 311-0193 (Japan); Tanaka, Teruya [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Tanigawa, Hisashi [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Okano, Kunihiko [Keio University, Yokohama, Kanagawa 223-0061 (Japan); Tobita, Kenji [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan); Kaneko, Osamu [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Ushigusa, Kenkichi [Rokkasho Fusion Institute, Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212 (Japan)

2016-11-01

Highlights: • The strategy for DEMO has been discussed by a joint effort in Japan. • DEMO should be aimed at steady power generation beyond several hundred MW. • DEMO should be aimed at availability extendable to commercialization. • DEMO should be aimed at tritium breeding to fulfill self-sufficiency of fuels. • Related actions are emerging to deliberate the Japanese fusion roadmap. - Abstract: The establishment of technology bases required for the development of a fusion demonstration reactor (DEMO) has been discussed by a joint effort throughout the Japanese fusion community. The basic concept of DEMO premised for investigation has been identified and the structure of technological issues to ensure the feasibility of this DEMO concept has been examined. The Joint-Core Team, which was launched along with the request by the ministerial council, has compiled analyses in two reports to clarify technology which should be secured, maintained, and developed in Japan, to share the common targets among industry, government, and academia, and to activate actions under a framework for implementation throughout Japan. The reports have pointed out that DEMO should be aimed at steady power generation beyond several hundred thousand kilowatts, availability which must be extended to commercialization, and overall tritium breeding to fulfill self-sufficiency of fuels. The necessary technological activities, such as superconducting coils, blanket, divertor, and others, have been sorted out and arranged in the chart with the time line toward the decision on DEMO. Based upon these Joint-Core Team reports, related actions are emerging to deliberate the Japanese fusion roadmap.

Dynamic modelling of balance of plant systems for a pulsed DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Harrington, C., E-mail: Chris.Harrington@ccfe.ac.uk

2015-10-15

Highlights: • A fully dynamic model of the balance of plant systems for pulsed DEMO is presented. • An operating strategy for handling pulse/dwell transitions has been devised. • Operation of a water-cooled system without energy storage appears feasible. • Steam turbine cycling can be minimised if rotation speed is maintained. - Abstract: The current baseline concept for a European DEMO defines a pulsed reactor producing power for periods of 2–4 h at a time, interrupted by dwell periods of approximately half an hour, potentially leading to cyclic fatigue of the heat transfer system and power generation equipment. Thermal energy storage systems could mitigate pulsing issues; however, the requirements for such a system cannot be defined without first understanding the challenges for pulsed operation, while any system will simultaneously increase the cost and complexity of the balance of plant. This work therefore presents a dynamic model of the primary heat transfer system and associated steam plant for a water-cooled DEMO, without energy storage, capable of simulating pulsed plant operation. An operating regime is defined such that the primary coolant flows continuously throughout the dwell period while the secondary steam flow is reduced. Simulation results show minimised thermal and pressure transients in the primary circuit, and small thermally induced stresses on the steam turbine rotor. If the turbine can be kept spinning to also minimise mechanical cycling, pulsed operation of a water-cooled DEMO without thermal energy storage may be feasible.

Methodology for Scaling Fusion Power Plant Availability

International Nuclear Information System (INIS)

Waganer, Lester M.

2011-01-01

Normally in the U.S. fusion power plant conceptual design studies, the development of the plant availability and the plant capital and operating costs makes the implicit assumption that the plant is a 10th of a kind fusion power plant. This is in keeping with the DOE guidelines published in the 1970s, the PNL report1, 'Fusion Reactor Design Studies - Standard Accounts for Cost Estimates. This assumption specifically defines the level of the industry and technology maturity and eliminates the need to define the necessary research and development efforts and costs to construct a one of a kind or the first of a kind power plant. It also assumes all the 'teething' problems have been solved and the plant can operate in the manner intended. The plant availability analysis assumes all maintenance actions have been refined and optimized by the operation of the prior nine or so plants. The actions are defined to be as quick and efficient as possible. This study will present a methodology to enable estimation of the availability of the one of a kind (one OAK) plant or first of a kind (1st OAK) plant. To clarify, one of the OAK facilities might be the pilot plant or the demo plant that is prototypical of the next generation power plant, but it is not a full-scale fusion power plant with all fully validated 'mature' subsystems. The first OAK facility is truly the first commercial plant of a common design that represents the next generation plant design. However, its subsystems, maintenance equipment and procedures will continue to be refined to achieve the goals for the 10th OAK power plant.

On the EU approach for DEMO architecture exploration and dealing with uncertainties

Energy Technology Data Exchange (ETDEWEB)

Coleman, M., E-mail: matti.coleman@euro-fusion.org [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Maviglia, F.; Bachmann, C. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); Anthony, J. [CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Federici, G. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); Shannon, M. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Wenninger, R. [EUROfusion Consortium, Boltzmannstraße 2, 85748 Garching (Germany); Max-Planck-Institut für Plasmaphysik, 85748 Garching (Germany)

2016-11-01

Highlights: • The issue of epistemic uncertainties in the DEMO design basis is described. • An approach to tackle uncertainty by investigating plant architectures is proposed. • The first wall heat load uncertainty is addressed following the proposed approach. - Abstract: One of the difficulties inherent in designing a future fusion reactor is dealing with uncertainty. As the major step between ITER and the commercial exploitation of nuclear fusion energy, DEMO will have to address many challenges – the natures of which are still not fully known. Unlike fission reactors, fusion reactors suffer from the intrinsic complexity of the tokamak (numerous interdependent system parameters) and from the dependence of plasma physics on scale – prohibiting design exploration founded on incremental progression and small-scale experimentation. For DEMO, this means that significant technical uncertainties will exist for some time to come, and a systems engineering design exploration approach must be developed to explore the reactor architecture when faced with these uncertainties. Important uncertainties in the context of fusion reactor design are discussed and a strategy for dealing with these is presented, treating the uncertainty in the first wall loads as an example.

Reduced activation structural materials for fusion power plants - The European Union program

International Nuclear Information System (INIS)

Schaaf, B. van der; Le Marois, G.; Moeslang, A.; Victoria, M.

2003-01-01

The competition of fusion power plants with the renewable energy sources in the second half of the 21st century requires structural materials operating at high temperatures, and sufficient radiation resistance to ensure high plant efficiency and availability. The reduced activation materials development in the EU counts several steps regarding the radiation damage resistance: 75 dpa for DEMO and 150 dpa and beyond for power plants. The maximum operating temperature development line ranges from the present day from the present day feasible 600 K up to 1300- K in advanced power plants. The reduced activation steel, RAS, forms the reference for the development efforts. EUROFER has been manufactured in the EU on industrial scale with specified purity and mechanical properties up to 825 K. The oxide dispersion strengthened , ODS, variety of RAS should reach the 925 K operation limit. The EU has selected silicon carbide ceramic composite as the primary high temperature, 1300 K, goal. On a small scale the potential of tungsten alloys for higher temperatures is investigated. The present test environments for radiation resistance are insufficient to provide data for DEMO. Hence the support of the EU for the International Fusion Materials Irradiation facility. The computational modelling is expected to guide the materials development and the design of near plasma components. The EU co-operates closely with Japan, the RF and US in IEA and IAEA co-ordinated agreements, which are highly beneficial for the fusion structural materials development. (author)

Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Di Gironimo, G. [ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2015-10-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon-Woo; Lee, Jeong-Hun [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University 1 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148, Yuseong-gu, Daejeon 305-806 (Korea, Republic of)

2016-02-15

Highlights: • A methodology to simulate the K-DEMO blanket system was proposed. • The results were compared with the CFD, to verify the prediction capability of MARS. • 46 Blankets in a single sector in K-DEMO were simulated using MARS-KS. • Supervisor program was devised to handle each blanket module individually. • The calculation results showed the flow rates, pressure drops, and temperatures. - Abstract: According to the conceptual design of the fusion DEMO reactor proposed by the National Fusion Research Institute of Korea, the water-cooled breeding blanket system incorporates a total of 736 blanket modules. The heat flux and neutron wall loading to each blanket module vary along their poloidal direction, and hence, thermal analysis for at least one blanket sector is required to confirm that the temperature limitations of the materials are satisfied in all the blanket modules. The present paper proposes a methodology of thermal analysis for multiple modules of the blanket system using a nuclear reactor thermal-hydraulic analysis code, MARS-KS. In order to overcome the limitations of the code, caused by the restriction on the number of computational nodes, a supervisor program was devised, which handles each blanket module separately at first, and then corrects the flow rate, considering pressure drops that occur in each module. For a feasibility test of the proposed methodology, 46 blankets in a single sector were simulated; the calculation results of the parameters, such as mass flow, pressure drops, and temperature distribution in the multiple blanket modules showed that the multi-module analysis method can be used for efficient thermal-hydraulic analysis of the fusion DEMO reactor.

Concept of DT fuel cycle for a fusion neutron source DEMO-FNS

Energy Technology Data Exchange (ETDEWEB)

Ananyev, Sergey S., E-mail: Ananyev_SS@nrcki.ru; Spitsyn, Alexander V.; Kuteev, Boris V.

2016-11-01

Highlights: • We presented the concept of a deuterium-tritium fuel cycle of stationary thermonuclear reactor. • Data of fuel cycles for nuclear facility (DEMO-FNS) with 2 variants of the fuel mixture for NBI system are presented. • The amount of tritium which is required for operation of DEMO-FNS is estimated. - Abstract: The paper describes the concept of a deuterium-tritium fuel cycle of a steady-state thermonuclear reactor with a fusion power over 10 MW. Parameters of fuel cycle for nuclear facility (JET scale) with different types of fuel mixtures for neutral beam injection system are presented. Optimization of fuel cycle characteristics was aimed at reducing flows and inventory of hydrogen isotopes and tritium in fuel cycle subsystems. The calculations were carried out using computer code TC-FNS to estimate tritium distribution in fusion reactor systems and components of “tritium plant”. The code enables calculations of tritium flows and inventory in the tokamak systems. Calculations of tritium flows and accumulation have been carried out for two different cases of the fuel mixture for neutral beam injection (NBI) system. The amounts of tritium which is required for operation of all fuel cycle systems in two different cases of the fuel mixture for NBI are 0.45 “” kg (D:T = 1:0) and 0.9 kg (D:T = 1:1) respectively.

Fuel cycle design for ITER and its extrapolation to DEMO

International Nuclear Information System (INIS)

Konishi, Satoshi; Glugla, Manfred; Hayashi, Takumi

2008-01-01

ITER is the first fusion device that continuously processes DT plasma exhaust and supplies recycled fuel in a closed loop. All the tritium and deuterium in the exhaust are recovered, purified and returned to the tokamak with minimal delay, so that extended burn can be sustained with limited inventory. To maintain the safety of the entire facility, plant scale detritiation systems will also continuously run to remove tritium from the effluents at the maximum efficiency. In this entire tritium plant system, extremely high decontamination factor, that is the ratio of the tritium loss to the processing flow rate, is required for fuel economy and minimized tritium emissions, and the system design based on the state-of-the-art technology is expected to satisfy all the requirements without significant technical challenges. Considerable part of the fusion tritium system will be verified with ITER and its decades of operation experiences. Toward the DEMO plant that will actually generate energy and operate its closed fuel cycle, breeding blanket and power train that caries high temperature and pressure media from the fusion device to the generation system will be the major addition. For the tritium confinement, safety and environmental emission, particularly blanket, its coolant, and generation systems such as heat exchanger, steam generator and turbine will be the critical systems, because the tritium permeation from the breeder and handling large amount of high temperature, high pressure coolant will be further more difficult than that required for ITER. Detritiation of solid waste such as used blanket and divertor will be another issue for both tritium economy and safety. Unlike in the case of ITER that is regarded as experimental facility, DEMO will be expected to demonstrate the safety, reliability and social acceptance issue, even if economical feature is excluded. Fuel and environmental issue to be tested in the DEMO will determine the viability of the fusion as a

Fuel cycle design for ITER and its extrapolation to DEMO

Energy Technology Data Exchange (ETDEWEB)

Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Kyoto 611-0011 (Japan)], E-mail: s-konishi@iae.kyoto-u.ac.jp; Glugla, Manfred [Forschungszentrum Karlsruhe, P.O. Box 3640, D 76021 Karlsruhe (Germany); Hayashi, Takumi [Apan Atomic Energy AgencyTokai, Ibaraki 319-0015 Japan (Japan)

2008-12-15

ITER is the first fusion device that continuously processes DT plasma exhaust and supplies recycled fuel in a closed loop. All the tritium and deuterium in the exhaust are recovered, purified and returned to the tokamak with minimal delay, so that extended burn can be sustained with limited inventory. To maintain the safety of the entire facility, plant scale detritiation systems will also continuously run to remove tritium from the effluents at the maximum efficiency. In this entire tritium plant system, extremely high decontamination factor, that is the ratio of the tritium loss to the processing flow rate, is required for fuel economy and minimized tritium emissions, and the system design based on the state-of-the-art technology is expected to satisfy all the requirements without significant technical challenges. Considerable part of the fusion tritium system will be verified with ITER and its decades of operation experiences. Toward the DEMO plant that will actually generate energy and operate its closed fuel cycle, breeding blanket and power train that caries high temperature and pressure media from the fusion device to the generation system will be the major addition. For the tritium confinement, safety and environmental emission, particularly blanket, its coolant, and generation systems such as heat exchanger, steam generator and turbine will be the critical systems, because the tritium permeation from the breeder and handling large amount of high temperature, high pressure coolant will be further more difficult than that required for ITER. Detritiation of solid waste such as used blanket and divertor will be another issue for both tritium economy and safety. Unlike in the case of ITER that is regarded as experimental facility, DEMO will be expected to demonstrate the safety, reliability and social acceptance issue, even if economical feature is excluded. Fuel and environmental issue to be tested in the DEMO will determine the viability of the fusion as a

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

Revised assessments of the economics of fusion power

International Nuclear Information System (INIS)

Han, W.E.; Ward, D.J.

2009-01-01

Although fusion power is being developed because of its large resource base, low environmental impact and high levels of intrinsic safety, it is also important to investigate the economics of a future fusion power plant in order to assess the potential market for the electricity produced. As part of the PPCS (Power Plant Conceptual Study) in Europe, published in 2005, an assessment was made of the likely economic performance of the range of fusion power plant concepts studied. Since that time, new work has been carried out, within the fusion programme, and particularly in the EU DEMO study, that changes a number of the important assumptions made in the PPCS. These changes allow either reduced cost versions of the PPCS plant models or, alternatively, plants with less ambitious technical assumptions at constant cost. The impact of the new results, emerging from the EU DEMO studies, on the role of fusion in the future energy market is described. A new energy economics model is employed to analyse the potential market performance of fusion power in a range of future energy scenarios and this shows that there can be a significant role for fusion in a future energy market.

Neutronic performance issues of the breeding blanket options for the European DEMO fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Bachmann, C. [EUROfusion—Programme Management Unit, Boltzmannstr. 2, 85748 Garching (Germany); Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, SERMA, LPEC, 91191 Gif-sur-Yvette (France); Moro, F. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Villari, R. [ENEA, Dipartimento Fusione e tecnologie per la Sicurezza Nucleare, Via E. Fermi 45, 00044 Frascati, Rome (Italy)

2016-11-01

Highlights: • Breeder blanket concepts for DEMO—design features. • Neutronic characteristics of breeder blankets. • Evaluation of Tritium breeding potential. • Evaluation of shielding performance. - Abstract: This paper presents nuclear performance issues of the HCPB, HCLL, DCLL and WCLL breeder blankets, which are under development within the PPPT (Power Plant Physics and Technology) programme of EUROfusion, with the objective to assess the potential and suitability of the blankets for the application to DEMO. The assessment is based on the initial design versions of the blankets developed in 2014. The Tritium breeding potential is considered sufficient for all breeder blankets although the initial design versions of the HCPB, HCLL and DCLL blankets were shown to require further design improvements. Suitable measures have been proposed and proven to be sufficient to achieve the required Tritium Breeding Ratio (TBR) ≥ 1.10. The shielding performance was shown to be sufficient to protect the super-conducting toroidal field coil provided that efficient shielding material mixtures including WC or borated water are utilized. The WCLL blanket does not require the use of such shielding materials due to a very compact blanket support structure/manifold configuration which yet requires design verification. The vacuum vessel can be safely operated over the full anticipated DEMO lifetime of 6 full power years for all blanket concepts considered.

Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade

DEFF Research Database (Denmark)

Meyer, H.; Abel, I.G.; Akers, R.J.

2013-01-01

New diagnostic, modelling and plant capability on the Mega Ampère Spherical Tokamak (MAST) have delivered important results in key areas for ITER/DEMO and the upcoming MAST Upgrade, a step towards future ST devices on the path to fusion currently under procurement. Micro-stability analysis...

DEMO diagnostics and burn control

Energy Technology Data Exchange (ETDEWEB)

Biel, Wolfgang, E-mail: w.biel@fz-juelich.de [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, Jülich (Germany); Department of Applied Physics, Ghent University (Belgium); Baar, Marco de [FOM-Institute DIFFER, Nieuwegein (Netherlands); Eindhoven University of Technology (Netherlands); Dinklage, Andreas [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Felici, Federico [Eindhoven University of Technology (Netherlands); König, Ralf [Max-Planck-Institut für Plasmaphysik, Greifswald (Germany); Meister, Hans; Treutterer, Wolfgang [Max-Planck-Institut für Plasmaphysik, Garching (Germany); Wenninger, Ronald [Max-Planck-Institut für Plasmaphysik, Garching (Germany); EFDA Power Plant Physics and Technology, Garching (Germany)

2015-10-15

Highlights: • An initial concept for the DEMO diagnostic and control system is presented. • A preliminary list of control functions and candidate diagnostics is developed. • Challenges regarding disruptions, power exhaust and radiation control are highlighted. • The need for introducing realistic control margins is emphasized. • On outline of the future R&D plan is presented. - Abstract: The development of the control system for a tokamak demonstration fusion reactor (DEMO) faces unprecedented challenges. First, the requirements for control reliability and accuracy are more stringent than on existing fusion devices: any loss of plasma control on DEMO may result in a disruption which could damage the inner wall of the machine, while operating the device with larger margins against the operational limits would lead to a reduction of the electrical output power. Second, the performance of DEMO control is limited by space restrictions for the implementation of components (optimization of the tritium breeding rate), by lifetime issues for the front-end parts (neutron and gamma radiation, erosion and deposition acting on all components) and by slow, weak and indirect action of the available actuators (plasma shaping, heating and fuelling). The European DEMO conceptual design studies include the development of a reliable control system, since the details of the achievable plasma scenario and the machine design may depend on the actual performance of the control system. In the first phase of development, an initial understanding of the prime choices of diagnostic methods applicable to DEMO, implementation and performance issues, the interrelation with the plasma scenario definition, and the planning of necessary future R&D have been obtained.

Study of the cooling systems with S-CO2 for the DEMO fusion power reactor.

Czech Academy of Sciences Publication Activity Database

Veselý, L.; Dostál, V.; Entler, Slavomír

2017-01-01

Roč. 124, November (2017), s. 244-247 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] Institutional support: RVO:61389021 Keywords : DEMO * Cooling * Energy conversion * Thermal cycle * Carbon dioxide * SCO2a Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617305719

Recent technical progress on BA Program: DEMO activities and IFMIF/EVEDA

Energy Technology Data Exchange (ETDEWEB)

Yamanishi, T.; Asakura, N.; Tobita, K.; Ohira, S. [Japan Atomic Energy Agency, Rokkasho, Aomori (Japan); Federici, G. [EFDA Close Support Unit, Garching (Germany); Heidinger, R. [Fusion for Energy, Garching (Germany); Knaster, J. [BA IFMIF/EVEDA Project Team, Rokkasho, Aomori (Japan); Clement, S. [Fusion for Energy, Barcelona (Spain); Nakajima, N. [BA IFERC Project Team, Rokkasho, Aomori (Japan)

2016-11-01

The Broader Approach (BA) activities consists of three major projects: the International Fusion Energy Research Center (IFERC) project, the International Fusion Materials Irradiation Facility/Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) project, and the Satellite Tokamak Programme (STP, JT-60SA). These projects have been carried out to obtain basic data for the design of DEMO fusion reactor from 2007. For 8-year activities, the above projects could produce a set of fruitful results for the DEMO reactor. DEMO design activity has been conducted to build a set of DEMO design bases in accordance with a series of discussion between EU and JA. In the DEMO R&D activities, five basic R&D subjects for a DEMO blanket system have been selected, and been studies under close collaborations between EU and JA: structure materials (RAFM steels and SiC/SiC composites), functional materials (tritium breeders and neutron multipliers), and tritium technology. Some additional R&D subjects recommended by peer review comments have also been studied successfully in recent years. Regarding the IFMIF/EVEDA project, some main components of the accelerator facility been designed and tested. The validation test using EVEDA Lithium Test Loop (ELTL) was also completed successfully in October 2014.

Maximal design basis accident of fusion neutron source DEMO-TIN

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

Technological and engineering challenges of fusion

International Nuclear Information System (INIS)

Maisonnier, David; Hayward, Jim

2008-01-01

The current fusion development scenario in Europe assumes the sequential achievement of key milestones. Firstly, the qualification of the DEMO/reactor physics basis in ITER, secondly, the qualification of materials for in-vessel components in IFMIF and, thirdly, the qualification of components and processes in DEMO. Although this scenario is constrained by budgetary considerations, it assumes the resolution of many challenges in physics, technology and engineering. In the first part of the paper, the technological and engineering challenges to be met in order to satisfy the current development scenario will be highlighted. These challenges will be met by an appropriate share of the work between ITER, IFMIF, DEMO and the necessary accompanying programme, which will have to include a number of dedicated facilities (e.g. for the development of H and CD systems). In the second part of the paper, the consequences of a considerable acceleration of the fusion development programme will be discussed. Although most of the technological and engineering challenges identified above will have to be met within a shorter timescale, it is possible to limit the requirements and expectation for a first fusion power plant with respect to those adopted for the current fusion development scenario. However, it must be recognised that such a strategy will inevitably result in increased risk and a reduction in the economy of the plant. (author)

Advances in the physics basis for the European DEMO design

Science.gov (United States)

Wenninger, R.; Arbeiter, F.; Aubert, J.; Aho-Mantila, L.; Albanese, R.; Ambrosino, R.; Angioni, C.; Artaud, J.-F.; Bernert, M.; Fable, E.; Fasoli, A.; Federici, G.; Garcia, J.; Giruzzi, G.; Jenko, F.; Maget, P.; Mattei, M.; Maviglia, F.; Poli, E.; Ramogida, G.; Reux, C.; Schneider, M.; Sieglin, B.; Villone, F.; Wischmeier, M.; Zohm, H.

2015-06-01

In the European fusion roadmap, ITER is followed by a demonstration fusion power reactor (DEMO), for which a conceptual design is under development. This paper reports the first results of a coherent effort to develop the relevant physics knowledge for that (DEMO Physics Basis), carried out by European experts. The program currently includes investigations in the areas of scenario modeling, transport, MHD, heating & current drive, fast particles, plasma wall interaction and disruptions.

Failure rate modeling using fault tree analysis and Bayesian network: DEMO pulsed operation turbine study case

International Nuclear Information System (INIS)

Dongiovanni, Danilo Nicola; Iesmantas, Tomas

2016-01-01

Highlights: • RAMI (Reliability, Availability, Maintainability and Inspectability) assessment of secondary heat transfer loop for a DEMO nuclear fusion plant. • Definition of a fault tree for a nuclear steam turbine operated in pulsed mode. • Turbine failure rate models update by mean of a Bayesian network reflecting the fault tree analysis in the considered scenario. • Sensitivity analysis on system availability performance. - Abstract: Availability will play an important role in the Demonstration Power Plant (DEMO) success from an economic and safety perspective. Availability performance is commonly assessed by Reliability Availability Maintainability Inspectability (RAMI) analysis, strongly relying on the accurate definition of system components failure modes (FM) and failure rates (FR). Little component experience is available in fusion application, therefore requiring the adaptation of literature FR to fusion plant operating conditions, which may differ in several aspects. As a possible solution to this problem, a new methodology to extrapolate/estimate components failure rate under different operating conditions is presented. The DEMO Balance of Plant nuclear steam turbine component operated in pulse mode is considered as study case. The methodology moves from the definition of a fault tree taking into account failure modes possibly enhanced by pulsed operation. The fault tree is then translated into a Bayesian network. A statistical model for the turbine system failure rate in terms of subcomponents’ FR is hence obtained, allowing for sensitivity analyses on the structured mixture of literature and unknown FR data for which plausible value intervals are investigated to assess their impact on the whole turbine system FR. Finally, the impact of resulting turbine system FR on plant availability is assessed exploiting a Reliability Block Diagram (RBD) model for a typical secondary cooling system implementing a Rankine cycle. Mean inherent availability

Failure rate modeling using fault tree analysis and Bayesian network: DEMO pulsed operation turbine study case

Energy Technology Data Exchange (ETDEWEB)

Dongiovanni, Danilo Nicola, E-mail: danilo.dongiovanni@enea.it [ENEA, Nuclear Fusion and Safety Technologies Department, via Enrico Fermi 45, Frascati 00040 (Italy); Iesmantas, Tomas [LEI, Breslaujos str. 3 Kaunas (Lithuania)

2016-11-01

Highlights: • RAMI (Reliability, Availability, Maintainability and Inspectability) assessment of secondary heat transfer loop for a DEMO nuclear fusion plant. • Definition of a fault tree for a nuclear steam turbine operated in pulsed mode. • Turbine failure rate models update by mean of a Bayesian network reflecting the fault tree analysis in the considered scenario. • Sensitivity analysis on system availability performance. - Abstract: Availability will play an important role in the Demonstration Power Plant (DEMO) success from an economic and safety perspective. Availability performance is commonly assessed by Reliability Availability Maintainability Inspectability (RAMI) analysis, strongly relying on the accurate definition of system components failure modes (FM) and failure rates (FR). Little component experience is available in fusion application, therefore requiring the adaptation of literature FR to fusion plant operating conditions, which may differ in several aspects. As a possible solution to this problem, a new methodology to extrapolate/estimate components failure rate under different operating conditions is presented. The DEMO Balance of Plant nuclear steam turbine component operated in pulse mode is considered as study case. The methodology moves from the definition of a fault tree taking into account failure modes possibly enhanced by pulsed operation. The fault tree is then translated into a Bayesian network. A statistical model for the turbine system failure rate in terms of subcomponents’ FR is hence obtained, allowing for sensitivity analyses on the structured mixture of literature and unknown FR data for which plausible value intervals are investigated to assess their impact on the whole turbine system FR. Finally, the impact of resulting turbine system FR on plant availability is assessed exploiting a Reliability Block Diagram (RBD) model for a typical secondary cooling system implementing a Rankine cycle. Mean inherent availability

Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Yanagi, Yoshihiko; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto; Kuroda, Toshimasa; Kosaku, Yasuo; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

2001-11-01

Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li{sub 2}O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo-critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m{sup 2}, {sup 6}Li enrichment of 30% and coolant temperature at inlet of breeder region of 380degC. In the case of the higher coolant temperature 430degC of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%. (author)

European blanket development for a demo reactor

International Nuclear Information System (INIS)

Giancarli, L.; Proust, E.; Anzidei, L.

1994-01-01

There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development. (authors). 8 figs., 1 tab

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)

Comparative study of cost models for tokamak DEMO fusion reactors

International Nuclear Information System (INIS)

Oishi, Tetsutarou; Yamazaki, Kozo; Arimoto, Hideki; Ban, Kanae; Kondo, Takuya; Tobita, Kenji; Goto, Takuya

2012-01-01

Cost evaluation analysis of the tokamak-type demonstration reactor DEMO using the PEC (physics-engineering-cost) system code is underway to establish a cost evaluation model for the DEMO reactor design. As a reference case, a DEMO reactor with reference to the SSTR (steady state tokamak reactor) was designed using PEC code. The calculated total capital cost was in the same order of that proposed previously in cost evaluation studies for the SSTR. Design parameter scanning analysis and multi regression analysis illustrated the effect of parameters on the total capital cost. The capital cost was predicted to be inside the range of several thousands of M$s in this study. (author)

European DEMO design strategy and consequences for materials

Science.gov (United States)

Federici, G.; Biel, W.; Gilbert, M. R.; Kemp, R.; Taylor, N.; Wenninger, R.

2017-09-01

Demonstrating the production of net electricity and operating with a closed fuel-cycle remain unarguably the crucial steps towards the exploitation of fusion power. These are the aims of a demonstration fusion reactor (DEMO) proposed to be built after ITER. This paper briefly describes the DEMO design options that are being considered in Europe for the current conceptual design studies as part of the Roadmap to Fusion Electricity Horizon 2020. These are not intended to represent fixed and exclusive design choices but rather ‘proxies’ of possible plant design options to be used to identify generic design/material issues that need to be resolved in future fusion reactor systems. The materials nuclear design requirements and the effects of radiation damage are briefly analysed with emphasis on a pulsed ‘low extrapolation’ system, which is being used for the initial design integration studies, based as far as possible on mature technologies and reliable regimes of operation (to be extrapolated from the ITER experience), and on the use of materials suitable for the expected level of neutron fluence. The main technical issues arising from the plasma and nuclear loads and the effects of radiation damage particularly on the structural and heat sink materials of the vessel and in-vessel components are critically discussed. The need to establish realistic target performance and a development schedule for near-term electricity production tends to favour more conservative technology choices. The readiness of the technical (physics and technology) assumptions that are being made is expected to be an important factor for the selection of the technical features of the device.

Preliminary analysis of K-DEMO thermal hydraulic system using MELCOR; Parametric study of hydrogen explosion

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Lim, Soo Min; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2016-10-15

K-DEMO (Korean fusion demonstration reactor) is future reactor for the commercializing the fusion power generation. The Design of K-DEMO is similar to that of ITER but the fusion energy generation is much bigger because ITER is experimental reactor. For this reason, K-DEMO uses more fusion reaction with bigger amount of tritium. Higher fusion power means more neutron generation that can irradiate the structure around fusion plasma. Fusion reactor can produce many kinds of radioactive material in the accident. Because of this hazard, preliminary safety analysis is mandatory before its construction. Concern for safety problem of accident of fusion/fission reactor has been growing after Fukushima accident which is severe accident from unexpected disaster. To model the primary heat transfer system, in this study, MARS-KS thermal hydraulic analysis is referred. Lee et al. and Kim et al. conducted thermal hydraulic analysis using MARS-KS and multiple module simulation to deal with the phenomena of first wall corrosion for each plasma pulse. This study shows the relationship between vacuum vessel rupture area and source term leakage after hydrogen explosion. For the conservative study, first wall heating is not terminated because the heating inside the vacuum vessel increase the pressure inside VV. Pressurizer, steam generator and turbine is not damaged. 6.69 kg of tritiated water (HTO) and 1 ton of dust is modeled which is ITER guideline. The entire system of K-DEMO is smaller than that of ITER. For this reason, lots of aerosol is release into environment although the safety system like DS is maintained. This result shows that the safety system of K-DEMO should use much more safety system.

Power plant conceptual studies in Europe

International Nuclear Information System (INIS)

Maisonnier, D.; Campbell, D.; Cook, I.; Pace, L. Di; Giancarli, L.; Hayward, J.; Puma, A. Li; Medrano, M.; Norajitra, P.; Roccella, M.; Sardain, P.; Tran, M.Q.; Ward, D.

2007-01-01

The European fusion programme is 'reactor oriented' and it is aimed at the successive demonstration of the scientific, the technological and the economic feasibility of fusion power. The European Power Plant Conceptual Study (PPCS) has been a study of conceptual designs of five commercial fusion power plants and the main emphasis was on system integration. It focused on five power plant models which are illustrative of a wider spectrum of possibilities. They are all based on the tokamak concept and they have approximately the same net electrical power output, 1500 MWe. These span a range from relatively near-term, based on limited technology and plasma physics extrapolations, to an advanced conception. The PPCS allows one to clarify the concept of DEMO, the device that will bridge the gap between ITER and the first-of-a-kind fusion power plant. An assessment of the PPCS models with limited extrapolations highlighted a number of issues that must be addressed to establish the DEMO physics and technological basis

Energy storage system for a pulsed DEMO

International Nuclear Information System (INIS)

Lucas, J.; Cortes, M.; Mendez, P.; Hayward, J.; Maisonnier, D.

2007-01-01

Several designs have been proposed for the DEMO fusion reactor. Some of them are working in a non-steady state mode. Since a power plant should be able to deliver to the grid a constant power, this challenge must be solved. Energy storage is required at a level of 250 MWh e with the capability of delivering a power of 1 GWe. A review of different technologies for energy storage is made. Thermal energy storage (TES), fuel cells and other hydrogen storage, compressed air storage, water pumping, batteries, flywheels and supercapacitors are the most promising solutions to energy storage. Each one is briefly described in the paper, showing its basis, features, advantages and disadvantages for this application. The conclusion of the review is that, based on existing technology, thermal energy storage using molten salts and a system based on hydrogen storage are the most promising candidates to meet the requirements of a pulsed DEMO. These systems are investigated in more detail together with an economic assessment of each

Towards a reduced activation structural materials database for fusion DEMO reactors

International Nuclear Information System (INIS)

Moeslang, A.; Diegele, E.; Laesser, R.; Klimiankou, M.; Lindau, R.; Materna-Morris, E.; Rieth, M.; Lucon, E.; Petersen, C.; Schneider, H.-C.; Pippan, R.; Rensman, J.W.; Schaaf, B. van der; Tavassoli, F.

2005-01-01

The development of First Wall, Blanket and Divertor materials which are capable of withstanding many years the high neutron and heat fluxes, is a critical path to fusion power. Therefore, the timely availability of a sound materials database has become an indispensable element in international fusion road maps. In order to provide materials design data for short term needs of ITER Test Blanket Modules and for a DEMOnstration fusion reactor, a wealth of R and D results on the European reduced activation ferritic-martensitic steel EUROFER, and on oxide dispersion strengthened variants are being characterized, mainly in the temperature window 250-650 deg. C. The characterisation includes irradiations up to 15 dpa in the mixed spectrum reactor HFR and up to 75 dpa in the fast breeder reactor BOR60. Industrial EUROFER-batches of 3.5 and 7.5 tons have been produced with a variety of semi-finished, quality-assured product forms. To increase thermal efficiency of blankets, high temperature resistant SiC f /SiC channel inserts for liquid metal coolant tubes are also developed. Regarding radiation damage resistance, a broad based reactor irradiation programs counts several steps from ≤5dpa (ITER TBMs) up to 75 dpa (DEMO). For the European divertor designers, a materials data base is presently being set up for pure W and W alloys, and related reactor irradiations are foreseen with temperatures from 650-1000 deg. C. (author)

Hydrogen production in early generation fusion power plant and its socio-economic implication

International Nuclear Information System (INIS)

Konishi, S.; Yamamoto, Y.

2007-01-01

Full text: This paper describes technical possibility of high temperature blanket for the early generation of fusion power plant and its application to hydrogen production. Its anticipated implication and strategy from the socio-economic aspects will be also discussed. Material and energy balances, such as fuel supply and delivery of product energy from fusion plants, as well as waste discharge and accident scenario that lead to environmental impact, are characterized by blanket concepts. Thus blankets are considered to dominate the feature of fusion energy that should respond to the requirements of the sponsors, i.e., public and future market. Fusion blanket concept based on the combinations of LiPb and SiC materials are regarded as a candidate for ITER/TBM, and at the same time, applied in various DEMO designs encompassing high temperature output. Recent developments of SiC-LiPb blanket in Japan, EU, US or China suggests staged development paths starting from TBMs and targeting high temperature blanket and efficient energy output from early generation plants. These strategies are strongly affected by the views of these parties on fusion energy, from the aspects of socio-economics. Hydrogen production process with the high temperature blanket is one of the most important issues, because temperature range much higher than is possible with current or near future fission plants are needed, suggesting market possibility different from that of fission. Fuel cycles, particularly lithium supply and TBR control will be also important. Self-sustained fusion fuel cycle requires technical capability to maintain the lithium contents. Liquid blanket has an advantage in continuous and real-time control TBR in a plant, but large amount of lithium-6 and initial tritium supply remains as issues. As for the environmental effect, normal operation release, assumed accidental scenario, and rad-waste will be the key issue to dominate social acceptance of fusion. (author)

Hydrogen production in early generation fusion power plant and its socio-economic implication

International Nuclear Information System (INIS)

Konishi, Satoshi; Yamamoto, Yasushi

2008-01-01

This paper describes technical possibility of high temperature blanket for the early generation of fusion power plant and its application to hydrogen production. Its anticipated implication and strategy from the socio-economic aspects will be also discussed. Material and energy balances, such as fuel supply and delivery of product energy from fusion plants, as well as waste discharge and accident scenario that lead to environmental impact, are characterized by blanket concepts. Thus blankets are considered to dominate the feature of fusion energy that should respond to the requirements of the sponsors, i.e., public and future market. Fusion blanket concept based on the combinations of LiPb and SiC materials are regarded as a candidate for ITER/TBM, and at the same time, applied in various DEMO designs encompassing high temperature output. Recent developments of SiC-LiPb blanket in Japan, EU, US or China suggests staged development paths starting from TBMs and targeting high temperature blanket and efficient energy output from early generation plants. These strategies are strongly affected by the views of these parties on fusion energy, from the aspects of socio-economics. Hydrogen production process with the high temperature blanket is one of the most important issues, because temperature range much higher than is possible with current or near future fission plants are needed, suggesting market possibility different from that of fission. Fuel cycles, particularly lithium supply and TBR control will be also important. Self-sustained fusion fuel cycle requires technical capability to maintain the lithium contents. Liquid blanket has an advantage in continuous and real-time control TBR in a plant, but large amount of lithium-6 and initial tritium supply remains as issues. As for the environmental effect, normal operation release, assumed accidental scenario, and rad-waste will be the key issue to dominate social acceptance of fusion. (author)

A roadmap to the realization of fusion energy

International Nuclear Information System (INIS)

Romanelli, Francesco

2013-01-01

With the reduction of CO2 emissions driving future energy policy, fusion can start market penetration beyond 2050 with up to 30% of electricity production by 2100. This requires an ambitious, yet realistic roadmap towards the demonstration of electricity production by 2050. This talk describes the main technical challenges on the path to fusion energy. For all of the challenges candidate solutions have been developed and the goal of the programme is now to demonstrate that they will also work at the scale of reactor. The roadmap has been developed within a goal-oriented approach articulated in eight different Missions. For each Mission the critical aspects for reactor application, the risks and risk mitigation strategies, the level of readiness now and after ITER and the gaps in the programme have been examined with involvement of experts from the ITER International Organization, Fusion for Energy, EFDA Close Support Unites and EFDA Associates. High-level work packages for the roadmap implementation have been prepared and the resources evaluated. ITER is the key facility in the roadmap and its success represents the most important overarching objectives of the EU programme. A demonstration fusion power plant (DEMO), producing net electricity for the grid at the level of a few hundreds MW is foreseen to start operation in the early 2040s. Following ITER, it will be the single step to a commercial fusion power plant. Industry must be involved early in the DEMO definition and design. The evolution of the programme requires that industry progressively shifts its role from that of provider of high-tech components to that of driver of the fusion development. Industry must be able to take full responsibility for the commercial fusion power plant after successful DEMO operation. For this reason, DEMO cannot be defined and designed by research laboratories alone, but requires the full involvement of industry in all technological and systems aspects of the design. Europe

Energy Storage System for a Pulsed DEMO

International Nuclear Information System (INIS)

Lucas, J.; Cortes, M.; Mendez, P.; Maisonnier, D.; Hayward, J.

2006-01-01

Several designs have been proposed for DEMO, some of which will operate in pulsed mode. Since a fusion power plant will be required to deliver continuous output, this challenge must be solved. For the reference DEMO, energy storage is required at a level of 250 MWhe with a capability of delivering a power of 1 GWe. Although DEMO is scheduled to be built in about 30 years, the design of the energy storage system must be based on current technology, focusing on commercially available products and on their expected future trends. From a thorough review of the different technologies available, thermal energy storage, compressed air energy storage, water pumping, fuel cells, batteries, flywheels and ultracapacitors are the most promising solutions to energy storage for a pulsed DEMO. An outline of each of these technologies is described in the paper, showing its basis, features, advantages and disadvantages for this application. Following this review, the most suitable methods capable of storing the required energy are examined. Fuel cells are not suitable due to the power requirement. Compressed air energy storage has a lower efficiency than the required one. Thermal energy storage, based on molten salts, so more energy can be stored with a better efficiency, and water pumping are shown as the main solutions, based on existing technology. However, those are not the only solutions capable of solving our challenge. Hydrogen production, using water electrolysis, hydrogen storage and combustion in a combined cycle can achieve our energy and power requirements with an acceptable efficiency. All these solutions are studied in detail and described, evaluating their current cost and efficiency in order to compare them all. (author)

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

Radioactive waste produced by DEMO and commerical fusion reactors extrapolated from ITER and advanced data bases

International Nuclear Information System (INIS)

Stacey, W.M.; Hertel, N.E.; Hoffman, E.A.

1994-01-01

The potential for providing energy with minimal environmental impact is a powerful motivation for the development of fusion and is the long-term objective of most fusion programs. However, the societal acceptability of magnetic fusion may well be decided in the near-term when decisions are taken on the construction of DEMO to follow ITER (if not when the construction decision is taken on ITER). Component wastes were calculated for DEMOs based on each data base by first calculating reactor sizes needed to satisfy the physics, stress and radiation attenuation requirements, and then calculating component replacement rates based on radiation damage and erosion limits. Then, radioactive inventories were calculated and compared to a number of international criteria for open-quote near-surface close-quote burial. None of the components in either type of design would meet the Japanese LLW criterion ( 3 ) within 10 years of shutdown, although the advanced (V/Li) blanket would do so soon afterwards. The vanadium first wall, divertor and blanket would satisfy the IAEA LLW criterion (<2 mSv/h contact dose) within about 10 years after shutdown, but none of the stainless steel or copper components would. All the components in the advanced data base designs except the stainless steel vacuum vessel and shield readily satisfy the US extended 10CFR61 intruder dose criterion, but none of the components in the open-quotes ITER data baseclose quotes designs do so. It seems unlikely that a stainless steel first wall or a copper divertor plate could satisfy the US (class C) criterion for near surface burial, much less the more stringent international, criteria. On the other hand, the first wall, divertor and blanket of the V/Li system would still satisfy the intruder dose concentration limits even if the dose criterion was reduced by two orders of magnitude

Objectives and status of EUROfusion DEMO blanket studies

Energy Technology Data Exchange (ETDEWEB)

Boccaccini, L.V., E-mail: lorenzo.boccaccini@kit.edu [Karlsruhe Institute of Technology (KIT) (Germany); Aiello, G.; Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Bachmann, C. [EUROfusion, PPPT, Garching (Germany); Barrett, T. [CCFE, Abingdon OX14 3DB (United Kingdom); Del Nevo, A. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Demange, D. [Karlsruhe Institute of Technology (KIT) (Germany); Forest, L. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-sur-Yvette (France); Hernandez, F.; Norajitra, P. [Karlsruhe Institute of Technology (KIT) (Germany); Porempovic, G. [Fuziotech Engineering Ltd (Hungary); Rapisarda, D. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Sardain, P. [CEA/IRFM, 13115 Saint-Paul-lès-Durance (France); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy); Vala, L. [Centrum výzkumu Řež, 250 68 Husinec-Řež (Czech Republic)

2016-11-01

Highlights: • Short description of the new Breeding Blanket Project in the EUROfusion consortium for the design of the EU PPPT DEMO: objectives. • Presentation of the design approach used in the development of the Breeding Blanket design: requirements. • Breeding Blanket design; in particular the four blanket concepts included in the study are presented, recent results highlighted and the status discussed. • Auxiliary systems and related R&D programme: in particular the work areas addressed in the Project (Tritium Technology, Pb-Li and Solid Breeders Technology, First Wall Design and R&D, Manufacturing) are presented, recent results highlighted and the status discussed. - Abstract: The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme.

Scientific and technical challenges on the road towards fusion electricity

Science.gov (United States)

Donné, A. J. H.; Federici, G.; Litaudon, X.; McDonald, D. C.

2017-10-01

The goal of the European Fusion Roadmap is to deliver fusion electricity to the grid early in the second half of this century. It breaks the quest for fusion energy into eight missions, and for each of them it describes a research and development programme to address all the open technical gaps in physics and technology and estimates the required resources. It points out the needs to intensify industrial involvement and to seek all opportunities for collaboration outside Europe. The roadmap covers three periods: the short term, which runs parallel to the European Research Framework Programme Horizon 2020, the medium term and the long term. ITER is the key facility of the roadmap as it is expected to achieve most of the important milestones on the path to fusion power. Thus, the vast majority of present resources are dedicated to ITER and its accompanying experiments. The medium term is focussed on taking ITER into operation and bringing it to full power, as well as on preparing the construction of a demonstration power plant DEMO, which will for the first time demonstrate fusion electricity to the grid around the middle of this century. Building and operating DEMO is the subject of the last roadmap phase: the long term. Clearly, the Fusion Roadmap is tightly connected to the ITER schedule. Three key milestones are the first operation of ITER, the start of the DT operation in ITER and reaching the full performance at which the thermal fusion power is 10 times the power put in to the plasma. The Engineering Design Activity of DEMO needs to start a few years after the first ITER plasma, while the start of the construction phase will be a few years after ITER reaches full performance. In this way ITER can give viable input to the design and development of DEMO. Because the neutron fluence in DEMO will be much higher than in ITER, it is important to develop and validate materials that can handle these very high neutron loads. For the testing of the materials, a

EURATOM strategy towards fusion energy

International Nuclear Information System (INIS)

Varandas, C.

2007-01-01

Research and development (Research and Development) activities in controlled thermonuclear fusion have been carried out since the 60's of the last century aiming at providing a new clean, powerful, practically inexhaustive, safe, environmentally friend and economically attractive energy source for the sustainable development of our society.The EURATOM Fusion Programme (EFP) has the leadership of the magnetic confinement Research and Development activities due to the excellent results obtained on JET and other specialized devices, such as ASDEX-Upgrade, TORE SUPRA, FTU, TCV, TEXTOR, CASTOR, ISTTOK, MAST, TJ-II, W7-X, RFX and EXTRAP. JET is the largest tokamak in operation and the single device that can use deuterium and tritium mixes. It has produced 16 MW of fusion power, during 3 seconds, with an energy amplification of 0.6. The next steps of the EFP strategy towards fusion energy are ITER complemented by a vigorous Accompanying Programme, DEMO and a prototype of a fusion power plant. ITER, the first experimental fusion reactor, is a large-scale project (35-year duration, 10000 MEuros budget), developed in the frame of a very broad international collaboration, involving EURATOM, Japan, Russia Federation, United States of America, Korea, China and India. ITER has two main objectives: (i) to prove the scientific and technical viability of fusion energy by producing 500 MW, during 300 seconds and a energy amplification between 10 and 20; and (ii) to test the simultaneous and integrated operation of the technologies needed for a fusion reactor. The Accompanying Programme aims to prepare the ITER scientific exploitation and the DEMO design, including the development of the International Fusion Materials Irradiation Facility (IFMIF). A substantial part of this programme will be carried out in the frame of the Broader Approach, an agreement signed by EURATOM and Japan. The main goal of DEMO is to produce electricity, during a long time, from nuclear fusion reactions. The

He-cooled divertor development for DEMO

International Nuclear Information System (INIS)

Norajitra, P.; Giniyatulin, R.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Kuznetsov, V.; Mazul, I.; Widak, V.; Ovchinnikov, I.; Ruprecht, R.; Zeep, B.

2007-01-01

Goal of the He-cooled divertor development for future fusion power plants is to resist a high heat flux of at least 10 MW/m 2 . The development includes the fields of design, analyses, and experiments. A helium-cooled modular jet concept (HEMJ) has been defined as reference solution, which is based on jet impingement cooling. In cooperation with the Efremov Institute, work was aimed at construction and high heat flux tests of prototypical tungsten mockups to demonstrate their manufacturability and their performances. A helium loop was built for this purpose to simulate the realistic thermo-hydraulics conditions close to those of DEMO (10 MPa He, 600 deg. C). The first high heat flux test results confirm the feasibility and the performance of the divertor design

Conceptuation of a continuously working vacuum pump train for fusion power plants

International Nuclear Information System (INIS)

Giegerich, Thomas; Day, Christian

2013-01-01

The Karlsruhe Institute of Technology (KIT) is developing a continuously working and non-cryogenic pumping solution for a demonstration power plant (DEMO). This pumping train shall cover the full operational pressure regime of a fusion reactor and is based on two pump types, namely diffusion pumps and liquid ring pumps. The whole pumping train must fulfill high safety and reliability requirements and it has to be made fully tritium compatible. In this paper, the design of a prototype pumping train and the special requirements for a DEMO machine are presented and discussed. A central feature of this pumping train is the use of a liquid metal as tritium compatible working fluid in both pump types, that leads to a pumping train which is able to cover a pressure range of 12 decades, namely from 10 −9 to 10 3 mbar. Finally, a test facility for pump testing over a wide pressure regime is described. In this facility (THESEUS), experiments with a diffusion pump have been performed and first results are presented

Conceptuation of a continuously working vacuum pump train for fusion power plants

Energy Technology Data Exchange (ETDEWEB)

Giegerich, Thomas, E-mail: thomas.giegerich@kit.edu; Day, Christian

2013-10-15

The Karlsruhe Institute of Technology (KIT) is developing a continuously working and non-cryogenic pumping solution for a demonstration power plant (DEMO). This pumping train shall cover the full operational pressure regime of a fusion reactor and is based on two pump types, namely diffusion pumps and liquid ring pumps. The whole pumping train must fulfill high safety and reliability requirements and it has to be made fully tritium compatible. In this paper, the design of a prototype pumping train and the special requirements for a DEMO machine are presented and discussed. A central feature of this pumping train is the use of a liquid metal as tritium compatible working fluid in both pump types, that leads to a pumping train which is able to cover a pressure range of 12 decades, namely from 10{sup −9} to 10{sup 3} mbar. Finally, a test facility for pump testing over a wide pressure regime is described. In this facility (THESEUS), experiments with a diffusion pump have been performed and first results are presented.

A Fast-Track Path to DEMO Enabled by ITER and FNSF-AT

Energy Technology Data Exchange (ETDEWEB)

Garofalo, A. M.; Choi, M.; Humphreys, D. A.; Kinsey, J. E.; Lao, L. L.; Snyder, P. B.; John, H. E.St.; Turnbull, A. D.; Taylor, T.S., E-mail: garofalo@fusion.gat.com [General Atomics, San Diego (United States); Chan, V. S.; Canik, J. M. [Oak Ridge National Laboratory, Oak Ridge (United States); Sawan, M. E. [University of Wisconsin, Madison (United States); Stangeby, P. C. [University of Toronto Institute for Aerospace Studies, Toronto (Canada)

2012-09-15

Full text: A Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) [1] is a key element of a fast track plan to a commercially attractive fusion DEMO. The next step forward on the path towards fusion commercialization must be a device that complements ITER in addressing the community identified science and technology gaps to DEMO, and that enables a DEMO construction decision triggered by the achievement of Q = 10 in ITER, presently scheduled for the year 2030. This paper elucidates the logic flow leading to the FNSF-AT approach for such a next step forward, and presents the results of recent analysis resolving key physics and engineering issues. A FNSF-AT will show fusion can make its own fuel, provide a materials irradiation facility, show fusion can produce high-grade process heat and electricity. In order to accomplish these goals, the FNSF has to operate steady-state with significant duty cycle and significant neutron fluence. In FNSF-AT, advanced tokamak physics enables steady-state burning plasmas with the high fluence required for FNSF's nuclear science development objective, in the compact size required to demonstrate Tritium fuel self-sufficiency using only a moderate quantity of the limited supply of Tritium. Physics based integrated modeling has found a steady-state baseline equilibrium with good stability and controllability properties. 2-D analysis assuming ITER heat and particle diffusion coefficients in the SOL predicts peak heat flux < 10 MW/m{sup 2} at the outer divertor targets. High fidelity and high-resolution 3D neutronics calculations have also been carried out, showing acceptable cumulative end-of-life organic insulator dose levels in all the device coils, and TBR > 1 for two blanket concepts considered. This FNSF-AT baseline plasma scenario has significant margin to meet the FNSF nuclear science mission. Moreover, the facility allows the development of more advanced scenarios to close the physics gaps to DEMO

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)

Neutronic analyses of the preliminary design of a DCLL blanket for the EUROfusion DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Palermo, Iole, E-mail: iole.palermo@ciemat.es; Fernández, Iván; Rapisarda, David; Ibarra, Angel

2016-11-01

Highlights: • We perform neutronic calculations for the preliminary DCLL Blanket design. • We study the tritium breeding capability of the reactor. • We determine the nuclear heating in the main components. • We verify if the shielding of the TF coil is maintained. - Abstract: In the frame of the newly established EUROfusion WPBB Project for the period 2014–2018, four breeding blanket options are being investigated to be used in the fusion power demonstration plant DEMO. CIEMAT is leading the development of the conceptual design of the Dual Coolant Lithium Lead, DCLL, breeding blanket. The primary role of the blanket is of energy extraction, tritium production, and radiation shielding. With this aim the DCLL uses LiPb as primary coolant, tritium breeder and neutron multiplier and Eurofer as structural material. Focusing on the achievement of the fundamental neutronic responses a preliminary blanket model has been designed. Thus detailed 3D neutronic models of the whole blanket modules have been generated, arranged in a specific DCLL segmentation and integrated in the generic DEMO model. The initial design has been studied to demonstrate its viability. Thus, the neutronic behaviour of the blanket and of the shield systems in terms of tritium breeding capabilities, power generation and shielding efficiency has been assessed in this paper. The results demonstrate that the primary nuclear performances are already satisfactory at this preliminary stage of the design, having obtained the tritium self-sufficiency and an adequate shielding.

Fusion for Energy: The European joint undertaking for ITER and the development of fusion energy

International Nuclear Information System (INIS)

Diegele, E.

2009-01-01

Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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)

Report on the diagnostics for control of the fusion DEMO reactors

International Nuclear Information System (INIS)

2014-05-01

The range of diagnostics that can be used in DEMO will be severely restricted compared to that used in the current experiments or to be used in ITER. Therefore, a study is planned on the technical feasibility of sensors and diagnostics on the basis of specific tokamak and helical DEMO designs, with the involvement of a wide range of specialists covering reactor design, diagnostics, neutronics, reactor structure, remote maintenance, plasma physics, plasma and machine control, and computer simulation. Topics included typical characteristic times of target plasma behavior, diagnostics tools with their resolution and lifetime, response time of actuators, and plasmas. Through these studies, possible candidates for DEMO diagnostics were identified. The outcome of two years of activities is summarized in this report with a recommendation to the government of Japan. (J.P.N.)

Updated conceptual design of helium cooling ceramic blanket for HCCB-DEMO

Energy Technology Data Exchange (ETDEWEB)

Wang, Suhao [University of Science and Technology of China, Hefei, Anhui (China); Southwestern Institute of Physics, Chengdu, Sichuan (China); Cao, Qixiang; Wu, Xinghua; Wang, Xiaoyu; Zhang, Guoshu [Southwestern Institute of Physics, Chengdu, Sichuan (China); Feng, Kaiming, E-mail: fengkm@swip.ac.cn [Southwestern Institute of Physics, Chengdu, Sichuan (China)

2016-11-15

Highlights: • An updated design of Helium Cooled Ceramic breeder Blanket (HCCB) for HCCB-DEMO is proposed in this paper. • The Breeder Unit is transformed to TBM-like sub-modules, with double “banana” shape tritium breeder. Each sub-module is inserted in space formed by Stiffen Grids (SGs). • The performance analysis is performed based on the R&D development of material, fabrication technology and safety assessment in CN ITER TBM program. • Hot spots will be located at the FW bend side. - Abstract: The basic definition of the HCCB-DEMO plant and preliminary blanket designed by Southwestern Institution of Physics was proposed in 2009. The DEMO fusion power is 2550 MW and electric power is 800 MW. Based on development of R&D in breeding blanket, a conceptual design of helium cooled blanket with ceramic breeder in HCCB-DEMO was presented. The main design features of the HCCB-DEMO blanket were: (1) CLF-1 structure materials, Be multiplier and Li{sub 4}SiO{sub 4} breeder; (2) neutronic wall load is 2.3 MW/m{sup 2} and surface heat flux is 0.43 MW/m{sup 2} (2) TBR ≈ 1.15; (3) geometry of breeding units is ITER TBM-like segmentation; (4)Pressure of helium is 8 MPa and inlet/outlet temperature is 300/500 °C. On the basis of these design, some important analytical results are presented in aspects of (i) neutronic behavior of the blanket; (ii) design of 3D structure and thermal-hydraulic lay-out for breeding blanket module; (iii) structural-mechanical behavior of the blanket under pressurization. All of these assessments proved current stucture fulfill the design requirements.

Diagnostics and control for the steady state and pulsed tokamak DEMO

Czech Academy of Sciences Publication Activity Database

Orsitto, F.P.; Villari, R.; Moro, F.; Todd, T.N.; Lilley, S.; Jenkins, I.; Felton, R.; Biel, W.; Silva, A.; Scholz, M.; Rzadkiewicz, J.; Ďuran, Ivan; Tardocchi, M.; Gorini, G.; Morlock, C.; Federici, G.; Litnovsky, A.

2016-01-01

Roč. 56, č. 2 (2016), č. článku 026009. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : measurement systems, fusion reactor, fusion plasma diagnostics * fusion reactor * fusion plasma diagnostics * DEMO * Hall sensors * tokamak Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/0029-5515/56/2/026009

A new meshless approach to map electromagnetic loads for FEM analysis on DEMO TF coil system

International Nuclear Information System (INIS)

Biancolini, Marco Evangelos; Brutti, Carlo; Giorgetti, Francesco; Muzzi, Luigi; Turtù, Simonetta; Anemona, Alessandro

2015-01-01

Graphical abstract: - Highlights: • Generation and mapping of magnetic load on DEMO using radial basis function. • Good agreement between RBF interpolation and EM TOSCA computations. • Resultant forces are stable with respect to the target mesh used. • Stress results are robust and accurate even if a coarse cloud is used for RBF interpolation. - Abstract: Demonstration fusion reactors (DEMO) are being envisaged to be able to produce commercial electrical power. The design of the DEMO magnets and of the constituting conductors is a crucial issue in the overall engineering design of such a large fusion machine. In the frame of the EU roadmap of the so-called fast track approach, mechanical studies of preliminary DEMO toroidal field (TF) coil system conceptual designs are being enforced. The magnetic field load acting on the DEMO TF coil conductor has to be evaluated as input in the FEM model mesh, in order to evaluate the stresses on the mechanical structure. To gain flexibility, a novel approach based on the meshless method of radial basis functions (RBF) has been implemented. The present paper describes this original and flexible approach for the generation and mapping of magnetic load on DEMO TF coil system.

The KALPUREX-process – A new vacuum pumping process for exhaust gases in fusion power plants

Energy Technology Data Exchange (ETDEWEB)

Giegerich, Thomas, E-mail: thomas.giegerich@kit.edu; Day, Christian

2014-10-15

Highlights: • A new vacuum pumping process for fusion power plants has been developed and is presented in this paper. • This process works continuously and non-cryogenic what leads to a strong reduction of the tritium inventory in the fuel cycle. • This pumping process is based on the use of a liquid metal (mercury) as working fluid and is called KALPUREX process. • The KALPUREX process is the technical realization of the DIR concept using a set of three vacuum pumps (metal foil pump/diffusion pump/liquid ring pump). • This paper discusses the arrangement of the pumps and also the required infrastructure for operation. - Abstract: The Karlsruhe Institute of Technology (KIT) is developing a continuously working and non-cryogenic pumping solution for torus exhaust pumping of a demonstration power plant (DEMO) including Direct Internal Recycling (DIR). This full pumping system consists of three pumps, namely a metal foil pump for gas separation, a linear diffusion pump as primary pump and a liquid ring pump as backing pump. The latter two pumps apply mercury as working fluid due to its perfect tritium compatibility. This asks for a baffle system on both sides of the pumping train to control working fluid vapour and to avoid any mercury propagation in the machine. In this paper, the arrangement of all torus pumps required for a power plant reactor as well as the corresponding infrastructure and its effect on the DEMO machine design is presented and discussed. The full pumping process is called ‘Karlsruhe liquid metal based pumping process for fusion reactor exhaust gases’ (KALPUREX process, patent pending)

The KALPUREX-process – A new vacuum pumping process for exhaust gases in fusion power plants

International Nuclear Information System (INIS)

Giegerich, Thomas; Day, Christian

2014-01-01

Highlights: • A new vacuum pumping process for fusion power plants has been developed and is presented in this paper. • This process works continuously and non-cryogenic what leads to a strong reduction of the tritium inventory in the fuel cycle. • This pumping process is based on the use of a liquid metal (mercury) as working fluid and is called KALPUREX process. • The KALPUREX process is the technical realization of the DIR concept using a set of three vacuum pumps (metal foil pump/diffusion pump/liquid ring pump). • This paper discusses the arrangement of the pumps and also the required infrastructure for operation. - Abstract: The Karlsruhe Institute of Technology (KIT) is developing a continuously working and non-cryogenic pumping solution for torus exhaust pumping of a demonstration power plant (DEMO) including Direct Internal Recycling (DIR). This full pumping system consists of three pumps, namely a metal foil pump for gas separation, a linear diffusion pump as primary pump and a liquid ring pump as backing pump. The latter two pumps apply mercury as working fluid due to its perfect tritium compatibility. This asks for a baffle system on both sides of the pumping train to control working fluid vapour and to avoid any mercury propagation in the machine. In this paper, the arrangement of all torus pumps required for a power plant reactor as well as the corresponding infrastructure and its effect on the DEMO machine design is presented and discussed. The full pumping process is called ‘Karlsruhe liquid metal based pumping process for fusion reactor exhaust gases’ (KALPUREX process, patent pending)

Fusion Technology for ITER, the ITER Project. Further Development Towards a DEMO Fusion Power Plant (3/4)

CERN Multimedia

CERN. Geneva

2011-01-01

This is the second half of a lecture series on fusion and will concentrate on fusion technology. The early phase of fusion development was concentrated on physics. However, during the 1980s it was realized that if one wanted to enter the area of fusion reactor plasmas, even in an experimental machine, a significant advance in fusion technologies would be needed. After several conceptual studies of reactor class fusion devices in the 1980s the engineering design phase of ITER started in earnest during the 1990s. The design team was in the beginning confronted with many challenges in the fusion technology area as well as in physics for which no readily available solution existed and in a few cases it was thought that solutions may be impossible to find. However, after the initial 3 years of intensive design and R&D work in an international framework utilizing basic fusion technology R&D from the previous decade it became clear that for all problems a conceptual solution could be found and further devel...

Preliminary analysis of the efficiency of non-standard divertor configurations in DEMO

Directory of Open Access Journals (Sweden)

F. Subba

2017-08-01

Full Text Available The standard Single Null (SN divertor is currently expected to be installed in DEMO. However, a number of alternative configurations are being evaluated in parallel as backup solutions, in case the standard divertor does not extrapolate successfully from ITER to a fusion power plant. We used the SOLPS code to produce a preliminary analysis of two such configurations, the X-Divertor (XD and the Super X-Divertor (SX, and compare them to the SN solution. Considering the nominal power flowing into the SOL (PSOL = 150 MW, we estimated the amplitude of the acceptable DEMO operational space. The acceptability criterion was chosen as plasma temperature at the target lower than 5eV, providing low sputtering and at least partial detachment, while the operational space was defined in terms of the electron density at the outboard mid-plane separatrix and of the seeded impurity (Ar only in the present study concentration. It was found that both the XD and the SXD extend the DEMO operational space, although the advantages detected so far are not dramatic. The most promising configuration seems to be the XD, which can produce acceptable target temperatures at moderate outboard mid-plane electron density (nomp=4.5×1019 m−3 and Zeff= 1.3.

Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

International Nuclear Information System (INIS)

Palermo, I.; Gómez-Ros, J.M.; Veredas, G.; Sanz, J.; Sedano, L.

2012-01-01

Highlights: ► Dual-Coolant He/Pb15.7Li breeding blanket for a DEMO fusion reactor is studied. ► An iterative process optimizes neutronic responses minimizing reactor dimension. ► A 3D toroidally symmetric geometry has been generated from the CAD model. ► Overall TBR values support the feasibility of the conceptual model considered. ► Power density in TF coils is below load limit for quenching. - Abstract: The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.

Thermal-hydraulic analysis of LTS cables for the DEMO TF coil using simplified models

Directory of Open Access Journals (Sweden)

Lewandowska Monika

2017-03-01

Full Text Available The conceptual design activities for the DEMOnstration reactor (DEMO – the prototype fusion power plant – are conducted in Europe by the EUROfusion Consortium. In 2015, three design concepts of the DEMO toroidal field (TF coil were proposed by Swiss Plasma Center (EPFL-SPC, PSI Villigen, Italian National Agency for New Technologies (ENEA Frascati, and Atomic Energy and Alternative Energies Commission (CEA Cadarache. The proposed conductor designs were subjected to complete mechanical, electromagnetic, and thermal-hydraulic analyses. The present study is focused on the thermal-hydraulic analysis of the candidate conductor designs using simplified models. It includes (a hydraulic analysis, (b heat removal analysis, and (c assessment of the maximum temperature and the maximum pressure in each conductor during quench. The performed analysis, aimed at verification whether the proposed design concepts fulfil the established acceptance criteria, provides the information for further improvements of the coil and conductors design.

Development, simulation and testing of structural materials for DEMO

International Nuclear Information System (INIS)

Laesser, R.; Baluc, N.; Boutard, J.-L.; Diegele, E.; Gasparotto, M.; Riccardi, B.; Dudarev, S.; Moeslang, A.; Pippan, R.; Schaaf, B. van der

2006-01-01

In DEMO the structural and functional materials of the in-vessel components will be exposed to a very intense flux of fusion neutrons with energies up to 14 MeV creating displacement cascades and gaseous transmutation products. Point defects and transmutations will induce new microstructures leading to changes in mechanical and physical properties such as hardening, swelling, loss of fracture toughness and creep strength. The kinetics of microstructural evolution depends on time, temperature and defect production rates. The structural materials to be used in DEMO should have very special properties: high radiation resistance up to the dose of 100 dpa, low residual activation, high creep strength and good compatibility with the cooling media in as wide a temperature operational window as possible for the achievement of high thermal efficiency. The most promising materials are: Reduced Activation Ferritic Martensitic (RAFM) steels (Eurofer and F82H), Oxide Dispersion Strengthened (ODS) RAFM and RAF steels, SiC fibres reinforced SiC matrix composites (SiCf/SiC), tungsten (W) and W-alloys. Each of these materials has its advantages and drawbacks and will be best used under certain conditions. Presently the best studied group of materials are the RAFM steels. They require the smallest extrapolation for use in DEMO but also offer the lowest upper temperature limit of operation (550 o C) and thus the lowest thermal efficiency. The other materials foreseen for more advanced breeder blanket and divertor concepts require intense fundamental R(and)D and testing before their acceptance, whereas the so-called Test Blanket Modules (TBMs) will be constructed using RAFM steel and tested in ITER. Validation of the DEMO structural materials will be done in IFMIF, the International Fusion Materials Irradiation Facility, which will produce neutron damage and transmutation products very similar to those characterising a fusion device and will allow accelerated testing with damage rates

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

Assessment of hypervapotron heat sink performance using CFD under DEMO relevant first wall conditions

Energy Technology Data Exchange (ETDEWEB)

Domalapally, Phani, E-mail: p_kumar.domalapally@cvrez.cz

2016-11-01

Highlights: • Performance of Hypervapotron heat sink was tested for First wall limiter application. • Two different materials were tested Eurofer 97 and CuCrZr at PWR conditions. • Simulations were performed to see the effect of the different inlet conditions and materials on the maximum temperature. • It was found that CuCrZr heat sink performance is far better than Eurofer heat sink at the same operating conditions. - Abstract: Among the proposed First Wall (FW) cooling concepts for European Demonstration Fusion Power Plant (DEMO), water cooled FW is one of the options. The heat flux load distribution on the FW of the DEMO reactor is not yet precisely defined. But if the heat loads on the FW are extrapolated from ITER conditions, the numbers are quite high and have to be handled none the less. The design of the FW itself is challenging as the thermal conductivity ratio of heat sink materials in ITER (CuCrZr) and in DEMO (Eurofer 97) is ∼10–12 and the operating conditions are of Pressurized Water Reactor (PWR) in DEMO instead of 70 °C and 4 MPa as in ITER. This paper analyzes the performance of Hypervapotron (HV) heat sink for FW limiter application under DEMO conditions. Where different materials, temperatures, heat fluxes and velocities are considered to predict the performance of the HV, to establish its limits in handling the heat loads before reaching the upper limits from temperature point of view. In order to assess the performance, numerical simulations are performed using commercial CFD code, which was previously validated in predicting the thermal hydraulic performance of HV geometry. Based on the results the potential usage of HV heat sink for DEMO will be assessed.

Metal Hall sensors for the new generation fusion reactors of DEMO scale

Science.gov (United States)

Bolshakova, I.; Bulavin, M.; Kargin, N.; Kost, Ya.; Kuech, T.; Kulikov, S.; Radishevskiy, M.; Shurygin, F.; Strikhanov, M.; Vasil'evskii, I.; Vasyliev, A.

2017-11-01

For the first time, the results of on-line testing of metal Hall sensors based on nano-thickness (50-70) nm gold films, which was conducted under irradiation by high-energy neutrons up to the high fluences of 1 · 1024 n · m-2, are presented. The testing has been carried out in the IBR-2 fast pulsed reactor in the neutron flux with the intensity of 1.5 · 1017 n · m-2 · s-1 at the Joint Institute for Nuclear Research. The energy spectrum of neutron flux was very close to that expected for the ex-vessel sensors locations in the ITER experimental reactor. The magnetic field sensitivity of the gold sensors was stable within the whole fluence range under research. Also, sensitivity values at the start and at the end of irradiation session were equal within the measurement error (<1%). The results obtained make it possible to recommend gold sensors for magnetic diagnostics in the new generation fusion reactors of DEMO scale.

The Logic-Based Supervisor Control for Sun-Tracking System of 1 MW HCPV Demo Plant: Study Case

Directory of Open Access Journals (Sweden)

Hong-Yih Yeh

2012-02-01

Full Text Available This paper presents a logic-based supervisor controller designed for trackers for a 1MW HCPV demo plant in Taiwan. A sun position sensor on the tracker is used to detect the sun position, as the sensor is sensitive to the intensity of sun light. The signal output of the sensor is partially affected by the cloud, which has a hard control position with the traditional PID control. Therefore we have used logic-based supervisor (LBS control which permits switching the PID control to sun trajectory under sunny or cloudy conditions. To verify the stability of the proposed control, an experiment was performed and the results show that the proposed control can efficiently achieve stabilization of the trackers of the 1MW HCPV demo plant.

Blanket/first wall challenges and required R&D on the pathway to DEMO

International Nuclear Information System (INIS)

Abdou, Mohamed; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

2015-01-01

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

Blanket/first wall challenges and required R&D on the pathway to DEMO

Energy Technology Data Exchange (ETDEWEB)

Abdou, Mohamed, E-mail: abdou@fusion.ucla.edu; Morley, Neil B.; Smolentsev, Sergey; Ying, Alice; Malang, Siegfried; Rowcliffe, Arthur; Ulrickson, Mike

2015-11-15

The breeding blanket with integrated first wall (FW) is the key nuclear component for power extraction, tritium fuel sustainability, and radiation shielding in fusion reactors. The ITER device will address plasma burn physics and plasma support technology, but it does not have a breeding blanket. Current activities to develop “roadmaps” for realizing fusion power recognize the blanket/FW as one of the principal remaining challenges. Therefore, a central element of the current planning activities is focused on the question: what are the research and major facilities required to develop the blanket/FW to a level which enables the design, construction and successful operation of a fusion DEMO? The principal challenges in the development of the blanket/FW are: (1) the Fusion Nuclear Environment – a multiple-field environment (neutrons, heat/particle fluxes, magnetic field, etc.) with high magnitudes and steep gradients and transients; (2) Nuclear Heating in a large volume with sharp gradients – the nuclear heating drives most blanket phenomena, but accurate simulation of this nuclear heating can be done only in a DT-plasma based facility; and (3) Complex Configuration with blanket/first wall/divertor inside the vacuum vessel – the consequence is low fault tolerance and long repair/replacement time. These blanket/FW development challenges result in critical consequences: (a) non-fusion facilities (laboratory experiments) need to be substantial to simulate multiple fields/multiple effects and must be accompanied by extensive modeling; (b) results from non-fusion facilities will be limited and will not fully resolve key technical issues. A DT-plasma based fusion nuclear science facility (FNSF) is required to perform “multiple effects” and “integrated” experiments in the fusion nuclear environment; and (c) the Reliability/Availability/Maintainability/Inspectability (RAMI) of fusion nuclear components is a major challenge and is one of the primary reasons

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

Energy Technology Data Exchange (ETDEWEB)

Merrill, B.J., E-mail: Brad.Merrill@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Wong, C.P.C. [General Atomics, San Diego, CA 92186-5608 (United States); Cadwallader, L.C. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID (United States); Abdou, M.; Morley, N.B. [University of California Los Angeles, Los Angeles, CA 90095-1597 (United States)

2014-10-15

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

Neutral beam deployment on DEMO and its influence on design

Energy Technology Data Exchange (ETDEWEB)

Surrey, Elizabeth, E-mail: elizabeth.surrey@ccfe.ac.uk [EURATOM/CCFE, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); King, Damian; Lister, Jonathan; Porton, Michael; Timmis, William; Ward, David [EURATOM/CCFE, Culham Science Centre, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)

2011-10-15

The demands on the neutral beam heating and current drive system of a DEMO device exceed those of existing fusion experiments by several orders of magnitude. By predicting possible power waveforms it is possible to analyse the technological advances necessary to achieve a system relevant to deployment on a power plant. Achieving the necessary efficiency will require simultaneous improvements in beam current density, neutralization efficiency and beam transmission. Considering the deployment on the tokamak vessel shows no major disruption to the tritium breeder blanket and no requirement to reach a high packing density of injectors. The thermal management of components subjected to low heat flux for many hours is considered and it is shown that radiation cooling can be exploited to control the temperature of such items.

Conceptual design studies of the Electron Cyclotron launcher for DEMO reactor

Science.gov (United States)

Moro, Alessandro; Bruschi, Alex; Franke, Thomas; Garavaglia, Saul; Granucci, Gustavo; Grossetti, Giovanni; Hizanidis, Kyriakos; Tigelis, Ioannis; Tran, Minh-Quang; Tsironis, Christos

2017-10-01

A demonstration fusion power plant (DEMO) producing electricity for the grid at the level of a few hundred megawatts is included in the European Roadmap [1]. The engineering design and R&D for the electron cyclotron (EC), ion cyclotron and neutral beam systems for the DEMO reactor is being performed by Work Package Heating and Current Drive (WPHCD) in the framework of EUROfusion Consortium activities. The EC target power to the plasma is about 50 MW, in which the required power for NTM control and burn control is included. EC launcher conceptual design studies are here presented, showing how the main design drivers of the system have been taken into account (physics requirements, reactor relevant operations, issues related to its integration as in-vessel components). Different options for the antenna are studied in a parameters space including a selection of frequencies, injection angles and launch points to get the best performances for the antenna configuration, using beam tracing calculations to evaluate plasma accessibility and deposited power. This conceptual design studies comes up with the identification of possible limits, constraints and critical issues, essential in the selection process of launcher setup solution.

Conceptual design studies of the Electron Cyclotron launcher for DEMO reactor

Directory of Open Access Journals (Sweden)

Moro Alessandro

2017-01-01

Full Text Available A demonstration fusion power plant (DEMO producing electricity for the grid at the level of a few hundred megawatts is included in the European Roadmap [1]. The engineering design and R&D for the electron cyclotron (EC, ion cyclotron and neutral beam systems for the DEMO reactor is being performed by Work Package Heating and Current Drive (WPHCD in the framework of EUROfusion Consortium activities. The EC target power to the plasma is about 50 MW, in which the required power for NTM control and burn control is included. EC launcher conceptual design studies are here presented, showing how the main design drivers of the system have been taken into account (physics requirements, reactor relevant operations, issues related to its integration as in-vessel components. Different options for the antenna are studied in a parameters space including a selection of frequencies, injection angles and launch points to get the best performances for the antenna configuration, using beam tracing calculations to evaluate plasma accessibility and deposited power. This conceptual design studies comes up with the identification of possible limits, constraints and critical issues, essential in the selection process of launcher setup solution.

Summary of the International Workshop on Magnetic Fusion Energy (MFE) Roadmapping in the ITER Era; 7–10 September 2011, Princeton, NJ, USA

International Nuclear Information System (INIS)

Neilson, G.H.; Federici, G.; Li, J.; Maisonnier, D.; Wolf, R.

2012-01-01

With the ITER project now well under way, the countries engaged in fusion research are planning, with renewed intensity, the research and major facilities needed to develop the science and technology for harnessing fusion energy. The Workshop on MFE Roadmapping in the ITER Era was organized to provide a timely forum for an international exchange of technical information and strategic perspectives on how best to tackle the remaining challenges leading to a magnetic fusion DEMO, a nuclear fusion device or devices with a level of physics and technology integration necessary to cover the essential elements of a commercial fusion power plant. Presentations addressed issues under four topics: (1) Perspectives on DEMO and the roadmap to DEMO; (2) Technology; (3) Physics-Technology integration and optimization; and (4) Major facilities on the path to DEMO. Participants identified a set of technical issues of high strategic importance, where the development strategy strongly influences the overall roadmap, and where there are divergent understandings in the world community, namely (1) the assumptions used in fusion design codes, (2) the strategy for fusion materials development, (3) the strategy for blanket development, (4) the strategy for plasma exhaust solution development and (5) the requirements and state of readiness for next-step facility options. It was concluded that there is a need to continue and to focus the international discussion concerning the scientific and technical issues that determine the fusion roadmap, and it was suggested that an international activity be organized under appropriate auspices to foster international cooperation on these issues. (conference report)

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

Energy Technology Data Exchange (ETDEWEB)

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N. [Institution Project center ITER, Moscow (Russian Federation)

2014-08-21

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

Science.gov (United States)

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

2014-08-01

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ-ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed.

Research at ITER towards DEMO: Specific reactor diagnostic studies to be carried out on ITER

International Nuclear Information System (INIS)

Krasilnikov, A. V.; Kaschuck, Y. A.; Vershkov, V. A.; Petrov, A. A.; Petrov, V. G.; Tugarinov, S. N.

2014-01-01

In ITER diagnostics will operate in the very hard radiation environment of fusion reactor. Extensive technology studies are carried out during development of the ITER diagnostics and procedures of their calibration and remote handling. Results of these studies and practical application of the developed diagnostics on ITER will provide the direct input to DEMO diagnostic development. The list of DEMO measurement requirements and diagnostics will be determined during ITER experiments on the bases of ITER plasma physics results and success of particular diagnostic application in reactor-like ITER plasma. Majority of ITER diagnostic already passed the conceptual design phase and represent the state of the art in fusion plasma diagnostic development. The number of related to DEMO results of ITER diagnostic studies such as design and prototype manufacture of: neutron and γ–ray diagnostics, neutral particle analyzers, optical spectroscopy including first mirror protection and cleaning technics, reflectometry, refractometry, tritium retention measurements etc. are discussed

Conceptual design study of the K-DEMO magnet system

Energy Technology Data Exchange (ETDEWEB)

Kim, Keeman, E-mail: kkeeman@nfri.re.kr [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Oh, Sangjun; Park, Jong Sung; Lee, Chulhee; Im, Kihak; Kim, Hyung Chan; Lee, Gyung-Su [National Fusion Research Institute, 169-148 Gwahak-ro, Daejeon 305-806 (Korea, Republic of); Neilson, George; Brown, Thomas; Kessel, Charles; Titus, Peter; Zhai, Yuhu [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

2015-10-15

Highlights: • Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. • Present a preliminary design of TF (toroidal field) magnet. • Present a preliminary design of CS (central solenoid) magnet. • Present a preliminary design of PF (toroidal field) magnet. - Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb{sub 3}Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

Conceptual design study of the K-DEMO magnet system

International Nuclear Information System (INIS)

Kim, Keeman; Oh, Sangjun; Park, Jong Sung; Lee, Chulhee; Im, Kihak; Kim, Hyung Chan; Lee, Gyung-Su; Neilson, George; Brown, Thomas; Kessel, Charles; Titus, Peter; Zhai, Yuhu

2015-01-01

Highlights: • Perform a preliminary conceptual study for a steady-state Korean DEMO reactor. • Present a preliminary design of TF (toroidal field) magnet. • Present a preliminary design of CS (central solenoid) magnet. • Present a preliminary design of PF (toroidal field) magnet. - Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb_3Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material.

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

Overview of Progress on the EU DEMO Reactor Magnet System Design

Czech Academy of Sciences Publication Activity Database

Zani, L.; Bayer, C.M.; Biancolini, M.E.; Bonifetto, R.; Bruzzone, P.; Brutti, C.; Ciazynski, D.; Coleman, M.; Ďuran, Ivan; Eisterer, M.; Fietz, W.H.; Gade, P.V.; Gaio, E.; Giorgetti, F.; Goldacker, W.; Gömöry, F.; Granados, X.; Heller, R.; Hertout, P.; Hoa, C.; Kario, A.; Lacroix, B.; Lewandowska, M.; Maistrello, A.; Muzzi, L.; Nijhuis, A.; Nunio, F.; Panin, A.; Petrisor, T.; Poncet, J.-M.; Prokopec, R.; Sanmarti Cardona, M.; Savoldi, L.; Schlachter, S.I.; Sedlak, K.; Stepanov, B.; Tiseanu, I.; Torre, A.; Turtu, S.; Vallcorba, R.; Vojenciak, M.; Weiss, K.-P.; Wesche, R.; Yagotintsev, K.; Zanino, R.

2016-01-01

Roč. 26, č. 4 (2016), č. článku 4204505. ISSN 1051-8223 Institutional support: RVO:61389021 Keywords : DEMO * fusion * HTS * LTS * Nb3Sn * superconducting magnets Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.092, year: 2015

Inertial fusion commercial power plants

International Nuclear Information System (INIS)

Logan, B.G.

1994-01-01

This presentation discusses the motivation for inertial fusion energy, a brief synopsis of five recently-completed inertial fusion power plant designs, some general conclusions drawn from these studies, and an example of an IFE hydrogen synfuel plant to suggest that future fusion studies consider broadening fusion use to low-emission fuels production as well as electricity

Development of a zonal applicability tool for remote handling equipment in DEMO

Energy Technology Data Exchange (ETDEWEB)

Madzharov, Vladimir, E-mail: vladimir.madzharov@kit.edu [Karlsruhe Institute of Technology, Institute for Material Handling and Logistics, Karlsruhe (Germany); Mittwollen, Martin [Karlsruhe Institute of Technology, Institute for Material Handling and Logistics, Karlsruhe (Germany); Leichtle, Dieter [Fusion for Energy F4E, Barcelona (Spain); Hermon, Gary [Culham Center for Fusion Energy, Culham Science Centre, OX14 3DB Abingdon (United Kingdom)

2015-10-15

Highlights: • Radiation-hardness assessment of remote handling (RH) components used in DEMO. • A radiation assessment tool for supporting remote handling engineers. • Connecting data from the radiation field analysis to the radiation hardness data. • Output is the expected lifetime of the selected RH component used for maintenance. - Abstract: A radiation-induced damage caused by the ionizing radiation can induce a malfunctioning of the remote handling equipment (RHE) used during maintenance in fusion power plants, other nuclear power stations and high-energy accelerators facilities like e.g. IFMIF. Therefore to achieve a sufficient length of operational time inside future fusion power plants, a suitable radiation tolerant RHE for maintenance operations in radiation environments is inevitably required. To assess the influence of the radiation on remote handling equipment (RHE), an investigation about radiation hardness assessment of typically used RHE components, has been performed. Additionally, information about the absorbed total dose that every component can withstand before failure was collected. Furthermore, the development of a zonal applicability tool for supporting RHE designers has been started using Excel VBA. The tool connects the data from the radiation field analysis (3-D radiation map) to the radiation hardness data of the planned RHE for DEMO remote maintenance. The intelligent combination of the available information for the radiation behaviour and radiation level at certain time and certain location may help with the taking of decisions about the application of RHE in radiation environment. The user inputs the following parameters: the specific device used in the RHE, the planned location and the maintenance period. The output is the expected lifetime of the selected RHE component at the given location and maintenance period. Planned action times have to be also considered. After having all the parameters it can be decided, if specific RHE

A preliminary systems assessment of the Starlite Demo candidates

International Nuclear Information System (INIS)

Bathke, C.G.

1995-01-01

The Starlite project has evaluated the following five tokamaks as candidates for the US Demo Power Plant: (1) steady state, first stability regime; (2) pulsed, first stability regime; (3) steady state, second stability regime; (4) steady state, reversed shear; and (5) steady state, low aspect ratio. Systems analysis of these candidates has played an important role in the selection of a reversed-shear tokamak for further conceptual design as a US Demo Power Plant. The cost-based systems analysis that led to the selection of a reversed-shear tokamak is described herein

Configuration management of the EU DEMO conceptual design data

Energy Technology Data Exchange (ETDEWEB)

Meszaros, Botond; Shannon, Mark [EUROfusion Consortium, PPPT Department, Garching, Boltzmannstr. 2 (Germany); Marzullo, Domenico [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Woodley, Colin; Rowe, Steve [CCFE, Culham Science Centre, Oxfordshire OX14 3DB, Abingdon (United Kingdom); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy)

2016-11-01

Highlights: • Description of the selection of the DEMO Product Data Management tool. • Introduction of the DEMO configuration management philosophy for the CAD design data. • Description of the enabling tools and systems of the configuration management. - Abstract: The EUROfusion Consortium is setting up – as part of the EU Fusion Roadmap – the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems “light weight” and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes.

Configuration management of the EU DEMO conceptual design data

International Nuclear Information System (INIS)

Meszaros, Botond; Shannon, Mark; Marzullo, Domenico; Woodley, Colin; Rowe, Steve; Di Gironimo, Giuseppe

2016-01-01

Highlights: • Description of the selection of the DEMO Product Data Management tool. • Introduction of the DEMO configuration management philosophy for the CAD design data. • Description of the enabling tools and systems of the configuration management. - Abstract: The EUROfusion Consortium is setting up – as part of the EU Fusion Roadmap – the framework for the implementation of the (pre)conceptual design phase of the DEMO reactor. Configuration management needs have been identified as one of the key elements of this framework and is the topic of this paper, in particular the configuration of the CAD design data. The desire is to keep the definition and layout of the corresponding systems “light weight” and relatively easy to manage, whilst simultaneously providing a level of detail in the definition of the design configuration that is fit for the purpose of a conceptual design. This paper aims to describe the steps followed during the definition of the configuration management system of the DEMO design data in terms of (i) the identification of the appropriate product data management system, (ii) the description of the philosophy of the configuration management of the design data, and (iii) the introduction of the most important enabling processes.

Creep behavior of 8Cr2WVTa martensitic steel designed for fusion DEMO reactor. An assessment on helium embrittlement resistance

International Nuclear Information System (INIS)

Yamamoto, Norikazu; Murase, Yoshiharu; Nagakawa, Johsei; Shiba, Kiyoyuki

2001-01-01

Mechanical response against transmutational helium production, alternatively susceptibility to helium embrittlement, in a nuclear fusion reactor was examined on 8Cr2WVTa martensitic steel, a prominent structural candidate for advanced fusion systems. In order to simulate DEMO (demonstrative) reactor environments, helium was implanted into the material at 823 K with concentrations up to 1000 appmHe utilizing an α-beam from a cyclotron. Creep rupture properties were subsequently determined at the same temperature and were compared with those of the material without helium. It has been proved that helium caused no meaningful deterioration in terms of both the creep lifetime and rupture elongation. Furthermore, failure occurred completely in a transgranular and ductile manner even after high concentration helium introduction and there was no symptom of grain boundary decohesion which very often arises in helium bearing materials. These facts would mirror preferable resistance of this steel toward helium embrittlement. (author)

Simulations with COREDIV Code of DEMO Discharges

Energy Technology Data Exchange (ETDEWEB)

Zagorski, R.; Stankiewicz, R.; Ivanova-Stanik, I., E-mail: roman.zagorski@ipplm.pl [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland)

2012-09-15

Full text: The reduction of divertor target power load due to radiation of sputtered and externally seeded impurities in fusion reactor is investigated in this paper. The approach is based on integrated numerical modelling of DEMO discharges using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multifluid transport in the SOL. The model is fully self-consistent with respect to both the effects of impurities on the alpha-power level and the interaction between seeded and intrinsic impurities. The code has been already successfully benchmarked with the data from present day experiments (JET, ADEX). Calculations have been performed for inductive DEMO scenario and DEMO Steady-State configuration with tungsten walls and Ar seeding. In case of DEMO Steady-State scenario strong increase of Z{sub eff} and significant reduction of the alpha power are observed with the increase of Ar influx which is caused by the decrease of fuel ions density due to the dilution effect. It leads to the reduction of the target plate heat loads but surprisingly the radiation level remains almost constant with the increased seeding which is the result of the interplay between the energy losses and tungsten source due to sputtering processes. It has been found that the W radiation is the dominant energy loss mechanism and it accounts for 90% of all radiation losses. In case of pulsed DEMO scenario, it appears that the helium accumulation might be a serious problem. Even without seeding the resulting Z{sub eff} is very large (> 2.6) and consequently only relatively weak seeding can be applied for pulsed scenario. It is found that helium accumulation depends strongly on the transport model used for helium, if the helium diffusion is increased than the accumulation effect is mitigated. (author)

European DEMO BOT solid breeder blanket

International Nuclear Information System (INIS)

Dalle Donne, M.

1994-11-01

The BOT (Breeder Outside Tube) Solid Breeder Blanket for a fusion DEMO reactor is presented. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. In the paper the reference blanket design and external loops are described as well as the results of the theoretical and experimental work in the fields of neutronics, thermohydraulics, mechanical stresses, tritium control and extraction, development and irradiation of the ceramic breeder material, beryllium development, ferromagnetic forces caused by disruptions, safety and reliability. An outlook is given on the remaining open questions and on the required R and D program. (orig.) [de

Demonstration tokamak-power-plant study (DEMO)

International Nuclear Information System (INIS)

1982-09-01

A study of a Demonstration Tokamak Power Plant (DEMO) has been completed. The study's objective was to develop a conceptual design of a prototype reactor which would precede commercial units. Emphasis has been placed on defining and analyzing key design issues and R and D needs in five areas: noninductive current drivers, impurity control systems, tritium breeding blankets, radiation shielding, and reactor configuration and maintenance features. The noninductive current drive analysis surveyed a wide range of candidates and selected relativistic electron beams for the reference reactor. The impurity control analysis considered both a single-null poloidal divertor and a pumped limiter. A pumped limiter located at the outer midplane was selected for the reference design because of greater engineering simplicity. The blanket design activity focused on two concepts: a Li 2 O solid breeder with high pressure water cooling and a lead-rich Li-Pb eutectic liquid metal breeder (17Li-83Pb). The reference blanket concept is the Li 2 O option with a PCA structural material. The first wall concept is a beryllium-clad corrugated panel design. The radiation shielding effort concentrated on reducing the cost of bulk and penetration shielding; the relatively low-cost outborad shield is composed of concrete, B 4 C, lead, and FE 1422 structural material

Fusion Implementation

International Nuclear Information System (INIS)

Schmidt, J.A.

2002-01-01

If a fusion DEMO reactor can be brought into operation during the first half of this century, fusion power production can have a significant impact on carbon dioxide production during the latter half of the century. An assessment of fusion implementation scenarios shows that the resource demands and waste production associated with these scenarios are manageable factors. If fusion is implemented during the latter half of this century it will be one element of a portfolio of (hopefully) carbon dioxide limiting sources of electrical power. It is time to assess the regional implications of fusion power implementation. An important attribute of fusion power is the wide range of possible regions of the country, or countries in the world, where power plants can be located. Unlike most renewable energy options, fusion energy will function within a local distribution system and not require costly, and difficult, long distance transmission systems. For example, the East Coast of the United States is a prime candidate for fusion power deployment by virtue of its distance from renewable energy sources. As fossil fuels become less and less available as an energy option, the transmission of energy across bodies of water will become very expensive. On a global scale, fusion power will be particularly attractive for regions separated from sources of renewable energy by oceans

DEMO relevance of the test blanket modules in ITER-Application to the European test blanket modules

International Nuclear Information System (INIS)

Magnani, E.; Gabriel, F.; Boccaccini, L.V.; Li-Puma, A.

2010-01-01

Test blanket module (TBM) testing programme in ITER as a support to DEMO design is a very important step on the road map to commercial fusion reactors although it is an ambitious task. Finding as much as possible DEMO relevant tests in view of the future DEMO blanket design is therefore a major goal since ITER and DEMO environment and loading conditions are different. To clarify and quantify the meaning of the DEMO relevance, criteria using a structural, functional and behavioural representation of the breeding blanket acting as a system are investigated. Then, a three-step strategy is proposed to carry out TBM DEMO relevant tests associated with a TBM design modification strategy. Key parameters should intensively be used as target for TBM characterization and numerical code validation. When assessing the relevance, on the other hand, not only the actual difference between DEMO and ITER values should be considered, but also whether the analyzed phenomena have a threshold and a range of applicability, as numerical simulations are usually permitted within these limits. The proposed methodology is at the end applied to the design of the HCLL TBM breeding unit configuration.

Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

International Nuclear Information System (INIS)

Ritz, G; Pintsuk, G; Linke, J; Hirai, T; Norajitra, P; Reiser, J; Giniyatulin, R; Makhankov, A; Mazul, I

2009-01-01

Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ∼14 MW m -2 , the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

Science.gov (United States)

Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

2009-12-01

Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

Cell fusion and nuclear fusion in plants.

Science.gov (United States)

Maruyama, Daisuke; Ohtsu, Mina; Higashiyama, Tetsuya

2016-12-01

Eukaryotic cells are surrounded by a plasma membrane and have a large nucleus containing the genomic DNA, which is enclosed by a nuclear envelope consisting of the outer and inner nuclear membranes. Although these membranes maintain the identity of cells, they sometimes fuse to each other, such as to produce a zygote during sexual reproduction or to give rise to other characteristically polyploid tissues. Recent studies have demonstrated that the mechanisms of plasma membrane or nuclear membrane fusion in plants are shared to some extent with those of yeasts and animals, despite the unique features of plant cells including thick cell walls and intercellular connections. Here, we summarize the key factors in the fusion of these membranes during plant reproduction, and also focus on "non-gametic cell fusion," which was thought to be rare in plant tissue, in which each cell is separated by a cell wall. Copyright © 2016 Elsevier Ltd. All rights reserved.

Next-Step Spherical Torus Experiment and Spherical Torus Strategy in the Fusion Energy Development Path

International Nuclear Information System (INIS)

Ono, M.; Peng, M.; Kessel, C.; Neumeyer, C.; Schmidt, J.; Chrzanowski, J.; Darrow, D.; Grisham, L.; Heitzenroeder, P.; Jarboe, T.; Jun, C.; Kaye, S.; Menard, J.; Raman, R.; Stevenson, T.; Viola, M.; Wilson, J.; Woolley, R.; Zatz, I.

2003-01-01

A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction

Overview of Progress on the EU DEMO Reactor Magnet System Design

NARCIS (Netherlands)

Zani, L.; Bayer, C.; biancolini, M.E.; Bonifetto, R.; Nijhuis, Arend; Yagotintsev, K.

2016-01-01

The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In

Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

Energy Technology Data Exchange (ETDEWEB)

Takatsu, H; Enoeda, M [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

1998-03-01

Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

Status on DEMO Helium Cooled Lithium Lead breeding blanket thermo-mechanical analyses

Energy Technology Data Exchange (ETDEWEB)

Aubert, J., E-mail: julien.aubert@cea.fr [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Aiello, G.; Jaboulay, J.-C. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Kiss, B. [Institute of Nuclear Techniques, Budapest University of Technology and Economics, Budapest (Hungary); Morin, A. [CEA-Saclay, DEN, DM2S, F-91191 Gif-sur-Yvette (France)

2016-11-01

Highlights: • CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. The DEMO HCLL breeding blanket design capitalizes on the experience acquired on the HCLL Test Blanket Module designed for ITER. Design improvements are being implemented to adapt the design to DEMO specifications and performance objectives. • Thermal and mechanical analyses have been carried out in order to justify the design of the HCLL breeding blanket showing promising results for tie rods modules’ attachments system and relatively good behavior of the box in case of LOCA when comparing to RCC-MRx criteria. • CFD thermal analyses on generic breeding unit have enabled the consolidation of the results obtained with previous FEM design analyses. - Abstract: The EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the breeding blanket surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. The Helium Cooled Lithium Lead (HCLL) blanket is one of the concepts which is investigated for DEMO. It is made of a Eurofer structure and uses the eutectic liquid lithium–lead as tritium breeder and neutron multiplier, and helium gas as coolant. Within the EUROfusion organization, CEA with the support of Wigner-RCP and IPP-CR, is in charge of the design of the HCLL blanket for DEMO. This paper presents the status of the thermal and mechanical analyses carried out on the HCLL breeding blanket in order to justify the design. CFD thermal analyses on generic breeding unit including stiffening plates and cooling plates have been performed with ANSYS in order to consolidate results obtained with previous FEM design analyses. Moreover in order to expand the justification of the HCLL Breeding blanket design, the most loaded area of

Pre-conceptual design study on K-DEMO ceramic breeder blanket

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Sung, E-mail: jspark@nfri.re.kr [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Kwon, Sungjin; Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Brown, Thomas; Neilson, George [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States)

2015-11-15

A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (B{sub T0} = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li{sub 4}SiO{sub 4} pebbles with Be{sub 12}Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials.

RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

International Nuclear Information System (INIS)

Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

2004-01-01

RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

Activation, decay heat, and waste classification studies of the European DEMO concept

Science.gov (United States)

Gilbert, M. R.; Eade, T.; Bachmann, C.; Fischer, U.; Taylor, N. P.

2017-04-01

Inventory calculations have a key role to play in designing future fusion power plants because, for a given irradiation field and material, they can predict the time evolution in chemical composition, activation, decay heat, gamma-dose, gas production, and even damage (dpa) dose. For conceptual designs of the European DEMO fusion reactor such calculations provide information about the neutron shielding requirements, maintenance schedules, and waste disposal prospects; thereby guiding future development. Extensive neutron-transport and inventory calculations have been performed for a reference DEMO reactor model with four different tritium-breeding blanket concepts. The results have been used to chart the post-operation variation in activity and decay heat from different vessel components, demonstrating that the shielding performance of the different blanket concepts—for a given blanket thickness—varies significantly. Detailed analyses of the simulated nuclide inventories for the vacuum vessel (VV) and divertor highlight the most dominant radionuclides, potentially suggesting how changes in material composition could help to reduce activity. Minor impurities in the raw composition of W used in divertor tiles, for example, are shown to produce undesirable long-lived radionuclides. Finally, waste classifications, based on UK regulations, and a recycling potential limit, have been applied to estimate the time-evolution in waste masses for both the entire vessel (including blanket modules, VV, divertor, and some ex-vessel components) and individual components, and also to suggest when a particular component might be suitable for recycling. The results indicate that the large mass of the VV will not be classifiable as low level waste on the 100 year timescale, but the majority of the divertor will be, and that both components will be potentially recyclable within that time.

Overview of materials research for fusion reactors

International Nuclear Information System (INIS)

Muroga, T.; Gasparotto, M.; Zinkle, S.J.

2002-01-01

Materials research for fusion reactors is overviewed from Japanese, EU and US perspectives. Emphasis is placed on programs and strategies for developing blanket structural materials, and recent highlights in research and development for reduced activation ferritic martensitic steels, vanadium alloys and SiC/SiC composites, and in mechanistic experimental and modeling studies. The common critical issue for the candidate materials is the effect of irradiation with helium production. For the qualification of materials up to the full lifetime of a DEMO and Power Plant reactors, an intense neutron source with relevant fusion neutron spectra is crucial. Elaborate use of the presently available irradiation devices will facilitate efficient and sound materials development within the required time scale

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.

Progress in the RAMI analysis of a conceptual LHCD system for DEMO

Science.gov (United States)

Mirizzi, F.

2014-02-01

Reliability, Availability, Maintainability and Inspectability (RAMI) concepts and techniques, that acquired great importance during the first manned space missions, have been progressively extended to industrial, scientific and consumer equipments to assure them satisfactory performances and lifetimes. In the design of experimental facilities, like tokamaks, mainly aimed at demonstrating validity and feasibility of scientific theories, RAMI analysis has been often left aside. DEMO, the future prototype fusion reactors, will be instead designed for steadily delivering electrical energy to commercial grids, so that the RAMI aspects will assume an absolute relevance since their initial design phases. A preliminary RAMI analysis of the LHCD system for the conceptual EU DEMO reactor is given in the paper.

Progress in the RAMI analysis of a conceptual LHCD system for DEMO

International Nuclear Information System (INIS)

Mirizzi, F.

2014-01-01

Reliability, Availability, Maintainability and Inspectability (RAMI) concepts and techniques, that acquired great importance during the first manned space missions, have been progressively extended to industrial, scientific and consumer equipments to assure them satisfactory performances and lifetimes. In the design of experimental facilities, like tokamaks, mainly aimed at demonstrating validity and feasibility of scientific theories, RAMI analysis has been often left aside. DEMO, the future prototype fusion reactors, will be instead designed for steadily delivering electrical energy to commercial grids, so that the RAMI aspects will assume an absolute relevance since their initial design phases. A preliminary RAMI analysis of the LHCD system for the conceptual EU DEMO reactor is given in the paper

Neutronics experiments for DEMO blanket at JAERI/FNS

International Nuclear Information System (INIS)

Sato, Satoshi; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

2003-01-01

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test speciments simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steels F82H were irradiated as typical fusion materials. The effective cross-sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

Neutronics experiments for DEMO blanket at JAERI/FNS

International Nuclear Information System (INIS)

Sato, S.; Ochiai, K.; Hori, J.; Verzilov, Y.; Klix, A.; Wada, M.; Terada, Y.; Yamauchi, M.; Morimoto, Y.; Nishitani, T.

2003-01-01

In order to verify the accuracy of the tritium production rate (TPR), neutron irradiation experiments have been performed with a mockup relevant to the fusion DEMO blanket consisting of F82H blocks, Li 2 TiO 3 blocks with a 6 Li enrichment of 40 and 95%, and beryllium blocks. Sample pellets of Li 2 TiO 3 were irradiated and the TPR was measured by a liquid scintillation counter. The TPR was also calculated using the Monte Carlo code MCNP-4B with the nuclear data library JENDL-3.2 and ENDF-B/VI. The results agreed with experimental values within the statistical error (10%) of the experiment. Accordingly, it was clarified that the TPR could be evaluated within 10% uncertainty by the calculation code and the nuclear data. In order to estimate the induced activity caused by sequential reactions in cooling water pipes in the DEMO blanket, neutron irradiation experiments have been performed using test specimens simulating the pipes. Sample metals of Fe, W, Ti, Pb, Cu, V and reduced activation ferritic steel F82H were irradiated as typical fusion materials. The effective cross- sections for incident neutron flux to calculate the radioactive nuclei ( 56 Co, 184 Re, 48 V, 206 Bi, 65 Zn and 51 Cr) due to sequential reactions were measured. From the experimental results, it was found that the effective cross-sections remarkably increases with coming closer to polyethylene board that was a substitute of water. As a result of the present study, it has become clear that the sequential reaction rates are important factors to accurately evaluate the induced activity in fusion reactors design. (author)

Steady State versus Pulsed Tokamak DEMO

Energy Technology Data Exchange (ETDEWEB)

Orsitto, F.P., E-mail: francesco.orsitto@enea.it [Associazione EURATOM-ENEA Unita Tecnica Fusione, Frascati (Italy); Todd, T. [CCFE/Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

2012-09-15

Full text: The present report deals with a Review of problems for a Steady state(SS) DEMO, related argument is treated about the models and the present status of comparison between the characteristics of DEMO pulsed versus a Steady state device.The studied SS DEMO Models (SLIM CS, PPCS model C EU-DEMO, ARIES-RS) are analyzed from the point of view of the similarity scaling laws and critical issues for a steady state DEMO. A comparison between steady state and pulsed DEMO is therefore carried out: in this context a new set of parameters for a pulsed (6 - 8 hours pulse) DEMO is determined working below the density limit, peak temperature of 20 keV, and requiring a modest improvement in the confinement factor(H{sub IPBy2} = 1.1) with respect to the H-mode. Both parameters density and confinement parameter are lower than the DEMO models presently considered. The concept of partially non-inductive pulsed DEMO is introduced since a pulsed DEMO needs heating and current drive tools for plasma stability and burn control. The change of the main parameter design for a DEMO working at high plasma peak temperatures T{sub e} {approx} 35 keV is analyzed: in this range the reactivity increases linearly with temperature, and a device with smaller major radius (R = 7.5 m) is compatible with high temperature. Increasing temperature is beneficial for current drive efficiency and heat load on divertor, being the synchrotron radiation one of the relevant components of the plasma emission at high temperatures and current drive efficiency increases with temperature. Technology and engineering problems are examined including efficiency and availability R&D issues for a high temperature DEMO. Fatigue and creep-fatigue effects of pulsed operations on pulsed DEMO components are considered in outline to define the R&D needed for DEMO development. (author)

Current fusion power plant design concepts

International Nuclear Information System (INIS)

Gore, B.F.; Murphy, E.S.

1976-09-01

Nine current U.S. designs for fusion power plants are described in this document. Summary tabulations include a tenth concept, for which the design document was unavailable during preparation of the descriptions. The information contained in the descriptions was used to define an envelope of fusion power plant characteristics which formed the basis for definition of reference first commercial fusion power plant design. A brief prose summary of primary plant features introduces each of the descriptions contained in the body of this document. In addition, summary tables are presented. These tables summarize in side-by-side fashion, plant parameters, processes, combinations of materials used, requirements for construction materials, requirements for replacement materials during operation, and production of wastes

Status of EC solid breeder blanket designs and R and D for demo fusion reactors

International Nuclear Information System (INIS)

Proust, E.; Anzidei, L.; Moons, F.

1994-01-01

Within the European Community Fusion Technology Program two solid breeder blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium purge gas, the martensitic steel MANET as structural material and beryllium as neutron multiplier. The configurations of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder materials are LiAlO 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li 4 SiO 4 and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behavior of the breeder ceramics and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation induced embrittlement of MANET, the mechanical effects caused by major plasma disruptions, and safety and reliability. The R and D work concentrate on these issues. The development of martensitic steels including MANET is part of a separate program. Breeder ceramics and beryllium irradiations have been so far performed for conditions which do not cover the peak values injected in the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. An effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling systems. First promising results have been obtained also in field of mechanical effects from plasma disruptions and safety and reliability, however further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects. (authors). 8 figs., 2 tabs., 53 refs

Exploration of one-dimensional plasma current density profile for K-DEMO steady-state operation

Energy Technology Data Exchange (ETDEWEB)

Kang, J.S. [Seoul National University, Seoul 151-742 (Korea, Republic of); Jung, L. [National Fusion Research Institute, Daejeon (Korea, Republic of); Byun, C.-S.; Na, D.H.; Na, Y.-S. [Seoul National University, Seoul 151-742 (Korea, Republic of); Hwang, Y.S., E-mail: yhwang@snu.ac.kr [Seoul National University, Seoul 151-742 (Korea, Republic of)

2016-11-01

Highlights: • One-dimensional current density and its optimization for the K-DEMO are explored. • Plasma current density profile is calculated with an integrated simulation code. • The impact of self and external heating profiles is considered self-consistently. • Current density is identified as a reference profile by minimizing heating power. - Abstract: Concept study for Korean demonstration fusion reactor (K-DEMO) is in progress, and basic design parameters are proposed by targeting high magnetic field operation with ITER-sized machine. High magnetic field operation is a favorable approach to enlarge relative plasma performance without increasing normalized beta or plasma current. Exploration of one-dimensional current density profile and its optimization process for the K-DEMO steady-state operation are reported in this paper. Numerical analysis is conducted with an integrated plasma simulation code package incorporating a transport code with equilibrium and current drive modules. Operation regimes are addressed with zero-dimensional system analysis. One-dimensional plasma current density profile is calculated based on equilibrium, bootstrap current analysis, and thermal transport analysis. The impact of self and external heating profiles on those parameters is considered self-consistently, where thermal power balance and 100% non-inductive current drive are the main constraints during the whole exploration procedure. Current and pressure profiles are identified as a reference steady-state profile by minimizing the external heating power with desired fusion power.

Net energy balance of tokamak fusion power plants

International Nuclear Information System (INIS)

Buende, R.

1981-10-01

The net energy balance for a tokamak fusion power plant was determined by using a PWR power plant as reference system, replacing the fission-specific components by fusion-specific components and adjusting the non-reactor-specific components to altered conditions. For determining the energy input to the fusion plant a method was developed that combines the advantages of the energetic input-output method with those of process chain analysis. A comparison with PWR, HTR, FBR, and coal-fired power plants is made. As a result the net energy balance of the fusion power plant turns out to be more advantageous than that of an LWR, HTR or coal-fired power plant and nearly in the same range as FBR power plants. (orig.)

Conceptual design studies for the European DEMO divertor: Rationale and first results

International Nuclear Information System (INIS)

You, J.H.; Mazzone, G.; Visca, E.; Bachmann, Ch.; Autissier, E.; Barrett, T.; Cocilovo, V.; Crescenzi, F.; Domalapally, P.K.; Dongiovanni, D.; Entler, S.; Federici, G.; Frosi, P.; Fursdon, M.; Greuner, H.; Hancock, D.; Marzullo, D.; McIntosh, S.; Müller, A.V.; Porfiri, M.T.

2016-01-01

Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

Conceptual design studies for the European DEMO divertor: Rationale and first results

Energy Technology Data Exchange (ETDEWEB)

You, J.H., E-mail: you@ipp.mpg.de [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Mazzone, G.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Bachmann, Ch. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Autissier, E. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Barrett, T. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Cocilovo, V.; Crescenzi, F. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Domalapally, P.K. [Research Cnter Rez, Hlavní 130, 250 68 Husinec–Řež (Czech Republic); Dongiovanni, D. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Entler, S. [Institute of Plasma Physics CAS, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Federici, G. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Frosi, P. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Fursdon, M. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Greuner, H. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Hancock, D. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Marzullo, D. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); McIntosh, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Müller, A.V. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Porfiri, M.T. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); and others

2016-11-01

Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

Net energy balance of tokamak fusion power plants

International Nuclear Information System (INIS)

Buende, R.

1983-01-01

The net energy balance for a tokamak fusion power plant of present day design is determined by using a PWR power plant as reference system, replacing the fission-specific components by fusion-specific components and adjusting the non-reactor-specific components to altered conditions. For determining the energy input to the fusion plant a method was developed that combines the advantages of the energetic input-output method with those of process chain analysis. A comparison with PWR, HTR, FBR, and coal-fired power plants is made. As a result the energy expenditures of the fusion power plant turn out to be lower than that of an LWR, HTR, or coal-fired power plant of equal net electric power output and nearly in the same range as FBR power plants. (orig.)

Overview of the design approach and prioritization of R&D activities towards an EU DEMO

Energy Technology Data Exchange (ETDEWEB)

Federici, G., E-mail: gianfranco.federici@euro-fusion.org [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Bachmann, C. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Biel, W. [Institute of Energy and Climate Research, Forschungszentrum Jülich GmbH, Jülich (Germany); Department of Applied Physics, Ghent University, Ghent (Belgium); Boccaccini, L. [Karlsruhe Institute of Technology (KIT), Campus Nord Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Cismondi, F.; Ciattaglia, S.; Coleman, M. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Day, C. [Karlsruhe Institute of Technology (KIT), Campus Nord Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Diegele, E.; Franke, T. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Grattarola, M. [Ansaldo Nucleare, Corso Perrone 25, 16152 Genova (Italy); Hurzlmeier, H. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Ibarra, A. [CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Loving, A. [CCFE Culham Science Centre, Abingdon OX14-3DB, Oxon (United Kingdom); Maviglia, F.; Meszaros, B.; Morlock, C. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Rieth, M. [Karlsruhe Institute of Technology (KIT), Campus Nord Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Shannon, M. [EUROfusion Consortium, Boltzmannstr. 2, Garching 85748 (Germany); Taylor, N. [CCFE Culham Science Centre, Abingdon OX14-3DB, Oxon (United Kingdom); and others

2016-11-01

Highlights: • An important objective of the EU fusion roadmap Horizon 2020 is to lay the foundation of a DEMO Fusion Power Reactor to follow ITER. • This paper describes the progress of the DEMO design and R&D activities in Europe in the EUROfusion Consortium. • Focus is on a systems engineering/design integration approach to identify technology & physics R&D requirements and address design challenges. • Preliminary design choices/sensitivity studies to explore the design space and identify/select attractive design points are described. • Initial results of work conducted by distributed project teams involving EU labs, universities, and industries in Europe are presented. - Abstract: This paper describes the progress of the DEMO design and R&D activities in Europe. The focus is on a systems engineering and design integration approach, which is recognized to be essential from an early stage to identify and address the engineering and operational challenges, and the requirements for technology and physics R&D. We present some of the preliminary design choices/sensitivity studies to explore and narrow down the design space and identify/select attractive design points. We also discuss some of the initial results of work being executed in the EUROfusion Consortium by a geographically distributed project team involving many EU laboratories, universities, and industries in Europe.

A new fully automatic PIM tool to replicate two component tungsten DEMO divertor parts

International Nuclear Information System (INIS)

Antusch, Steffen; Commin, Lorelei; Heneka, Jochen; Piotter, Volker; Plewa, Klaus; Walter, Heinz

2013-01-01

Highlights: • Development of a fully automatic 2C-PIM tool. • Replicate fusion relevant components in one step without additional brazing. • No cracks or gaps in the seam of the joining zone visible. • For both material combinations a solid bond of the material interface was achieved. • PIM is a powerful process for mass production as well as for joining even complex shaped parts. -- Abstract: At Karlsruhe Institute of Technology (KIT), divertor design concepts for future nuclear fusion power plants beyond ITER are intensively investigated. One promising KIT divertor design concept for the future DEMO power reactor is based on modular He-cooled finger units. The manufacturing of such parts by mechanical machining such as milling and turning, however, is extremely cost and time intensive because tungsten is very hard and brittle. Powder Injection Molding (PIM) has been adapted to tungsten processing at KIT since a couple of years. This production method is deemed promising in view of large-scale production of tungsten parts with high near-net-shape precision, hence, offering an advantage of cost-saving process compared to conventional machining. The properties of the effectively and successfully manufactured divertor part tile consisting only of pure tungsten are a microstructure without cracks and a high density (>98% T.D.). Based on the achieved results a new fully automatic multicomponent PIM tool was developed and allows the replication and joining without brazing of fusion relevant components of different materials in one step and the creation of composite materials. This contribution describes the process route to design and engineer a new fully automatic 2C-PIM tool, including the filling simulation and the implementing of the tool. The complete technological fabrication process of tungsten 2C-PIM, including material and feedstock (powder and binder) development, injection molding, and heat-treatment of real DEMO divertor parts is outlined

A high frequency, high power CARM proposal for the DEMO ECRH system

International Nuclear Information System (INIS)

Mirizzi, Francesco; Spassovsky, Ivan; Ceccuzzi, Silvio; Dattoli, Giuseppe; Di Palma, Emanuele; Doria, Andrea; Gallerano, Gianpiero; Lampasi, Alessandro; Maffia, Giuseppe; Ravera, GianLuca; Sabia, Elio; Tuccillo, Angelo Antonio; Zito, Pietro

2015-01-01

Highlights: • ECRH system for DEMO. • Cyclotron Auto-Resonance Maser (CARM) devices. • Relativistic electron beams. • Bragg reflectors. • High voltage pulse modulators. - Abstract: ECRH&CD systems are extensively used on tokamak plasmas due to their capability of highly tailored power deposition, allowing very localised heating and non-inductive current drive, useful for MHD and profiles control. The high electron temperatures expected in DEMO will require ECRH systems with operating frequency in the 200–300 GHz range, equipped with a reasonable number of high power (P ≥ 1 MW) CW RF sources, for allowing central RF power deposition. In this frame the ENEA Fusion Department (Frascati) is coordinating a task force aimed at the study and realisation of a suitable high power, high frequency reliable source.

A high frequency, high power CARM proposal for the DEMO ECRH system

Energy Technology Data Exchange (ETDEWEB)

Mirizzi, Francesco, E-mail: francesco.mirizzi@enea.it [Consorzio CREATE, Via Claudio 21, I-80125 Napoli (Italy); Spassovsky, Ivan [Unità Tecnica Applicazioni delle Radiazioni – ENEA, C.R. Frascati, via E. Fermi 45, I-00044 Frascati (Italy); Ceccuzzi, Silvio [Unità Tecnica Fusione – ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Dattoli, Giuseppe; Di Palma, Emanuele; Doria, Andrea; Gallerano, Gianpiero [Unità Tecnica Applicazioni delle Radiazioni – ENEA, C.R. Frascati, via E. Fermi 45, I-00044 Frascati (Italy); Lampasi, Alessandro; Maffia, Giuseppe; Ravera, GianLuca [Unità Tecnica Fusione – ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Sabia, Elio [Unità Tecnica Applicazioni delle Radiazioni – ENEA, C.R. Frascati, via E. Fermi 45, I-00044 Frascati (Italy); Tuccillo, Angelo Antonio; Zito, Pietro [Unità Tecnica Fusione – ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy)

2015-10-15

Highlights: • ECRH system for DEMO. • Cyclotron Auto-Resonance Maser (CARM) devices. • Relativistic electron beams. • Bragg reflectors. • High voltage pulse modulators. - Abstract: ECRH&CD systems are extensively used on tokamak plasmas due to their capability of highly tailored power deposition, allowing very localised heating and non-inductive current drive, useful for MHD and profiles control. The high electron temperatures expected in DEMO will require ECRH systems with operating frequency in the 200–300 GHz range, equipped with a reasonable number of high power (P ≥ 1 MW) CW RF sources, for allowing central RF power deposition. In this frame the ENEA Fusion Department (Frascati) is coordinating a task force aimed at the study and realisation of a suitable high power, high frequency reliable source.

Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

International Nuclear Information System (INIS)

Stankunas, Gediminas; Tidikas, Andrius; Pereslavstev, Pavel; Catalán, Juan; García, Raquel; Ogando, Francisco; Fischer, Ulrich

2016-01-01

Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

Activity inventories and decay heat calculations for a DEMO with HCPB and HCLL blanket modules

Energy Technology Data Exchange (ETDEWEB)

Stankunas, Gediminas, E-mail: gediminas.stankunas@lei.lt [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Tidikas, Andrius [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos Str. 3, LT-44403 Kaunas (Lithuania); Pereslavstev, Pavel [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Catalán, Juan; García, Raquel; Ogando, Francisco [Departamento de Ingeniería Energética, UNED, 28040 Madrid (Spain); Fischer, Ulrich [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2016-11-01

Highlights: • The afterheat and activity inventories were calculated for Eurofer steel which is the reference structural material for DEMO. • The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short and longer cooling times. • The comparison calculations were performed for a single outboard blanket module of the HCLL DEMO assuming High-Temperature Ferritic–Martensitic (HT-FM) steel and SS-316 (LN) as structural material. - Abstract: Activation inventories, decay heat and radiation doses are important nuclear quantities which need to be assessed on a reliable basis for the safe operation of a fusion nuclear power reactor. The afterheat and activity inventories were shown to be dominated by the Eurofer steel which is the reference structural material for DEMO. The decay heat for the HCPB DEMO was found to be larger than for the HCLL both for short (a few days) and longer (more than a year) cooling times. As for the alternative steels, the induced radioactivity was turned out to be lowest for the SS-316 until about 200 years after shut-down. Afterwards, the activity level of SS-316 steel was found to be the highest. For these times, the activity of both Eurofer and the HT-FM steel is about one order of magnitude lower.

Fusion power plant simulations: a progress report

International Nuclear Information System (INIS)

Cook, J.M.; Pattern, J.S.; Amend, W.E.

1976-01-01

The objective of the fusion systems analysis at ANL is to develop simulations to compare alternative conceptual designs of magnetically confined fusion power plants. The power plant computer simulation progress is described. Some system studies are also discussed

Integrated Computational study of Material Lifetime in a Fusion Reactor Environment

Energy Technology Data Exchange (ETDEWEB)

Gilbert, M.; Dudarev, S.; Packer, L.; Zheng, S.; Sublet, J.-C., E-mail: mark.gilbert@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Centre for Fusion Energy, Abingdon (United Kingdom)

2012-09-15

Full text: The high-energy, high-intensity neutron fluxes produced by the fusion plasma will have a significant life-limiting impact on reactor components in both experimental and commercial fusion devices. Not only do the neutrons bombarding the materials induce atomic displacement cascades, leading to the accumulation of structural defects, but they also initiate nuclear reactions, which cause transmutation of the elemental atoms. Understanding the implications associated with the resulting compositional changes is one of the key outstanding issues related to fusion energy research. Several complimentary computational techniques have been used to investigate the problem. Firstly, neutron-transport simulations, performed on a reference design for the demonstration fusion power plant (DEMO), quantify the variation in neutron irradiation conditions as a function of geometry. The resulting neutron fluxes and spectra are then used as input into inventory calculations, which allow for the compositional changes of a material to be tracked in time. These calculations reveal that the production of helium (He) gas atoms, whose presence in a material is of particular concern because it can accumulate and cause swelling and embrittlement, will vary significantly, even within the same component of a reactor. Lastly, a density-functional-based model for He-induced grain-boundary embrittlement has been developed to predict the life-limiting consequences associated with relatively low concentrations of He in materials situated at various locations in the DEMO structure. The results suggest that some important fusion materials may be significantly more susceptible to this type of failure than others. (author)

Fusion material development program in the broader approach activities

Energy Technology Data Exchange (ETDEWEB)

Nishitani, T. [Directorates of Fusion Energy Research: Naka, Ibaraki, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Tanigawa, H.; Jitsukawa, S. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Hayashi, K.; Takatsu, H. [Fusion Research and Development Directorate, Japan Momie Energy Agency, Ibaraki-ken (Japan); Yamanishi, T. [Tritium Process Laboratory, Japan Atomic Energy Research Institute, Tokai-mura, Ibaraki-ken (Japan); Tsuchiya, K. [Directorates of Fusion Energy Research, JAEA, Higashi-ibaraki-gun, Ibaraki-ken (Japan); MoIslang, A. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Baluc, N. [EPFL-Ecole Polytechnique Federale de Lausanne, Association Euratom-Confederation Suisse, UHD - CRPP, PPB, Lausanne (Switzerland); Pizzuto, A. [ENEA CR Frascat, Frascati (Italy); Hodgson, E.R. [CIEMAT-Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, Association Euratom-CIEMAT, Madrid (Spain); Lasser, R.; Gasparotto, M. [EFDA CSU Garching (Germany)

2007-07-01

Full text of publication follows: The world fusion community is now launching construction of ITER, the first nuclear-grade fusion machine in the world. In parallel to the ITER program, Broader Approach (BA) activities are initiated by EU and Japan, mainly at Rokkasho BA site in Japan. The BA activities include the International Fusion Materials Irradiation Facility-Engineering Validation and Engineering Design Activities (IFMIF-EVEDA), the International Fusion Energy Research Center (IFERC), and the Satellite Tokamak. IFERC consists of three sub project; a DEMO Design and R and D coordination Center, a Computational Simulation Center, and an ITER Remote Experimentation Center. Technical R and Ds mainly on fusion materials will be implemented as a part of the DEMO Design and R and D coordination Center. Based on the common interest of each party toward DEMO, R and Ds on a) reduced activation ferritic martensitic (RAFM) steels as a DEMO blanket structural material, SiCf/SiC composites, advanced tritium breeders and neutron multiplier for DEMO blankets, and Tritium Technology were selected and assessed by European and Japanese experts. In the R and D on the RAFM steels, the fabrication technology, techniques to incorporate the fracture/rupture properties of the irradiated materials, and methods to predict the deformation and fracture behaviors of structures under irradiation will be investigated. For SiCf/SiC composites, standard methods to evaluate high-temperature and life-time properties will be developed. Not only for SiCf/SiC but also related ceramics, physical and chemical properties such as He and H permeability and absorption will be investigated under irradiation. As the advanced tritium breeder R and D, Japan and EU plan to establish the production technique for advanced breeder pebbles of Li{sub 2}TiO{sub 3} and Li{sub 4}SiO{sub 4}, respectively. Also physical, chemical, and mechanical properties will be investigated for produced breeder pebbles. For the

The Direct Internal Recycling concept to simplify the fuel cycle of a fusion power plant

International Nuclear Information System (INIS)

Day, Christian; Giegerich, Thomas

2013-01-01

Highlights: • The fusion fuel cycle is presented and its functions are discussed. • Tritium inventories are estimated for an early DEMO configuration. • The Direct Internal Recycling concept to reduce tritium inventories is described. • Concepts for its technical implementation are developed. -- Abstract: A new concept, the Direct Internal Recycling (DIR) concept, is proposed, which minimizes fuel cycle inventory by adding an additional short-cut between the pumped torus exhaust gas and the fuelling systems. The paper highlights quantitative modelling results derived from a simple fuel cycle spreadsheet which underline the potential benefits that can be achieved by implementation of the DIR concept into a fusion power plant. DIR requires a novel set-up of the torus exhaust pumping system, which replaces the batch-wise and cyclic operated cryogenic pumps by a continuous pumping solution and which offers at the same time an additional integral gas separation function. By that, hydrogen can be removed close to the divertor from all other gases and the main load to the fuel clean-up systems is a smaller, helium-rich gas stream. Candidate DIR relevant pump technology based on liquid metals (vapour diffusion and liquid ring pumps) and metal foils is discussed

Fuel procurement for first generation fusion power plants

International Nuclear Information System (INIS)

Gore, B.F.; Hendrickson, P.L.

1976-09-01

The provision of deuterium, tritium, lithium and beryllium fuel materials for fusion power plants is examined in this document. Possible fusion reactions are discussed for use in first generation power plants. Requirements for fuel materials are considered. A range of expected annual consumption is given for each of the materials for a 1000 megawatts electric (MWe) fusion power plant. Inventory requirements are also given. Requirements for an assumed fusion power plant electrical generating capacity of 10 6 MWe (roughly twice present U.S. generating capacity) are also given. The supply industries are then examined for deuterium, lithium, and beryllium. Methods are discussed for producing the only tritium expected to be purchased by a commercial fusion industry--an initial inventory for the first plant. Present production levels and methods are described for deuterium, lithium and beryllium. The environmental impact associated with production of these materials is then discussed. The toxicity of beryllium is described, and methods are indicated to keep worker exposure to beryllium as low as achievable

Towards diagnostics for a fusion reactor

International Nuclear Information System (INIS)

Costley, A. E.

2009-01-01

The requirements for measurements on modern tokamak fusion plasmas are outlined, and the techniques and systems used to make the measurements, usually referred to as 'diagnostics', are introduced. The basics of three particular diagnostics - magnetics, neutron systems and a laser based optical system - are outlined as examples of modern diagnostic systems, and the implementation of these diagnostics on a current tokamak (JET) are described. The next major step in magnetic confinement fusion is the construction and operation of the International Thermonuclear Experimental Reactor (ITER), which is a joint project of China, Europe, Japan, India, Korea, the Russian Federation, and the United States. Construction has begun in Cadarache, France. It is expected that ITER will operate at the 500 MW level. Because of the harsh environment in the vacuum vessel where many diagnostic components are located, the development of diagnostics for ITER is a major challenge - arguably the most difficult challenge ever undertaken in the field of diagnostics. The main elements in the diagnostic step are outlined using the three chosen techniques as examples. Finally, the step beyond ITER to a demonstration reactor, DEMO, that is expected to produce several GWs of fusion power is considered and the impact on diagnostics outlined. It is shown that the applicability and development steps needed for the individual diagnostics techniques will differ. The challenges for DEMO diagnostics are substantial and a dedicated effort should be made to find and develop new techniques, and especially techniques appropriate to the DEMO environment. It is argued that the limitations and difficulties in diagnostics should be a consideration in the optimization and designs of DEMO. (author)

Self-consistent Analysis of a Blanket and Shielding of a Fusion Reactor Concept

International Nuclear Information System (INIS)

Kim, Suk Kwon; Hong, B. G.; Lee, D. W.; Kim, D. H.; Lee, Y. O.

2008-01-01

To develop the concept of a DEMO reactor, a tokamak reactor system analysis code has been developed at KAERI (Korea Atomic Energy Research Institute). The system analysis code incorporates prospects of the development of plasma physics and the technologies in a simple mathematical model and it helps to develop the concept of a fusion reactor and to identify the necessary R and D areas for a realization of the concept. In the system code, a plant power balance equation and a plasma power balance equation are solved to find plant parameters which satisfy the plasma physics and technology constraints, simultaneously. The outcome of the system analysis is to identify which areas of plasma physics and technologies and to what extent they should be developed for a realization of given fusion reactor concepts

Inertial Fusion Power Plant Concept of Operations and Maintenance

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-15

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

Overview of physics results from MAST towards ITER/DEMO and the MAST Upgrade

Czech Academy of Sciences Publication Activity Database

Meyer, H.; Abel, I.G.; Akers, R.J.; Allan, A.; Allan, S.Y.; Appel, L.C.; Asunta, O.; Barnes, M.; Barratt, N.C.; Ben Ayed, N.; Bradley, J.W.; Canik, J.; Cahyna, Pavel; Cecconelo, M.; Challis, C.D.; Chapman, I.T.; Ciric, D.; Colyer, G.; Conway, N.J.; Cox, M.; Crowley, B.J.; Cowley, S.C.; Cunningham, G.; Danilov, A.; Darke, A.; De Bock, M.F.M.; De Temmerman, G.; Dendy, R.O.; Denner, P.; Dickinson, D.; Dnestrovsky, A.Y.; Dnestrovsky, Y.; Driscoll, M.D.; Dudson, B.; Dunai, D.; Dunstan, M.; Dura, P.; Elmore, S.; Field, A.R.; Fishpool, G.; Freethy, S.; Fundameski, W.; Garzotti, L.; Ghim, Y.C.; Gibson, K.J.; Gryaznevich, M.P.; Harrison, J.; Havlíčková, E.; Hawkes, N.C.; Heidbrink, W.W.; Hender, T.C.; Highcock, E.; Higgins, D.; Hill, P.; Hnat, B.; Hole, M.J.; Horáček, Jan; Howell, D.F.; Imada, K.; Jones, O.; Kaveeva, E.; Keeling, D.; Kirk, A.; Kočan, M.; Lake, R.J.; Lehnen, M.; Leggate, H.J.; Liang, Y.; Lilley, M.K.; Lisgo, S.W.; Liu, Y.Q.; Lloyd, B.; Maddison, G.P.; Mailloux, J.; Martin, R.; McArdle, G.J.; McClements, K.G.; McMillan, B.; Michael, C.; Militello, F.; Molchanov, P.; Mordijck, S.; Morgan, T.; Morris, A.W.; Muir, D.G.; Nardon, E.; Naulin, V.; Naylor, G.; Nielsen, A.H.; O’Brien, M.R.; O’Gorman, T.; Pamela, S.; Parra, F.I.; Patel, A.; Pinches, S.D.; Price, M.N.; Roach, C.M.; Robinson, J.R.; Romanelli, M.; Rozhansky, V.; Saarelma, S.; Sangaroon, S.; Saveliev, A.; Scannell, R.; Seidl, J.; Sharapov, S.E.; Schekochihin, A.A.; Shevchenko, V.; Shibaev, S.; Stork, D.; Storrs, J.; Sykes, A.; Tallents, G. J.; Tamain, P.; Taylor, D.; Temple, D.; Thomas-Davies, N.; Thornton, A.; Turnyanskiy, M.R.; Valovič, M.; Vann, R.G.L.; Verwichte, E.; Voskoboynikov, P.; Voss, G.; Warder, S.E.V.; Wilson, H. R.; Wodniak, I.; Zoletnik, S.; Zagórski, R.

2013-01-01

Roč. 53, č. 10 (2013), s. 104008-104008 ISSN 0029-5515. [IAEA Fusion Energy Conference/24./. San Diego, 08.10.2012-13.10.2012] Institutional support: RVO:61389021 Keywords : ITER * DEMO * MAST * spherical tokamak * JET Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.243, year: 2013 http://iopscience.iop.org/0029-5515/53/10/104008/pdf/0029-5515_53_10_104008.pdf

Tore supra: towards the 'long time' fusion. Press journey

International Nuclear Information System (INIS)

2008-01-01

After a recall of the interest in the fusion for the development of energy sources, the document presents the fusion from the solar reaction to the nuclear fusion in laboratory. Then it discusses the great challenges of this technology and the Tore Supra installation. The last part is devoted to ITER and DEMO projects. (A.L.B.)

Tore supra: towards the 'long time' fusion. Press journey; Tore Supra: vers la fusion 'longue duree'. Voyage de presse

Energy Technology Data Exchange (ETDEWEB)

NONE

2008-07-01

After a recall of the interest in the fusion for the development of energy sources, the document presents the fusion from the solar reaction to the nuclear fusion in laboratory. Then it discusses the great challenges of this technology and the Tore Supra installation. The last part is devoted to ITER and DEMO projects. (A.L.B.)

Fusion advanced studies Torus

International Nuclear Information System (INIS)

2007-01-01

The successful development of ITER and DEMO scenarios requires preparatory activities on devices that are smaller than ITER, sufficiently flexible and capable of investigating the peculiar physics of burning plasma conditions. The aim of the Fusion Advanced Studies Torus (FAST) proposal [2.1] (formerly FT3 [2.2]) is to show that the preparation of ITER scenarios and the development of new expertise for the DEMO design and RD can be effectively implemented on a new facility. FAST will a) operate with deuterium plasmas, thereby avoiding problems associated with tritium, and allow investigation of nonlinear dynamics (which are important for understanding alpha particle behaviour in burning plasmas) by using fast ions accelerated by heating and current drive systems; b) work in a dimensionless parameter range close to that of ITER; c) test technical innovative solutions, such as full-tungsten plasma-facing components and an advanced liquid metal divertor target for the first wall/divertor, directly relevant for ITER and DEMO; d) exploit advanced regimes with a much longer pulse duration than the current diffusion time; e) provide a test bed for ITER and DEMO diagnostics; f) provide an ideal framework for model and numerical code benchmarks, their verification and validation in ITER/ DEMO-relevant plasma conditions

Fusion power plant economics

International Nuclear Information System (INIS)

Miller, R.L.

1996-01-01

The rationale, methodology, and updated comparative results of cost projections for magnetic-fusion-energy central-station electric power plants are considered. Changing market and regulatory conditions, particularly in the U.S., prompt fundamental reconsideration of what constitutes a competitive future energy-source technology and has implications for the direction and emphasis of appropriate near-term research and development programs, for fusion and other advanced generation systems. 36 refs., 2 figs., 2 tabs

Materials availability for fusion power plant construction

International Nuclear Information System (INIS)

Hartley, J.N.; Erickson, L.E.; Engel, R.L.; Foley, T.J.

1976-09-01

A preliminary assessment was made of the estimated total U.S. material usage with and without fusion power plants as well as the U.S. and foreign reserves and resources, and U.S. production capacity. The potential environmental impacts of fusion power plant material procurement were also reviewed including land alteration and resultant chemical releases. To provide a general measure for the impact of material procurement for fusion reactors, land requirements were estimated for mining and disposing of waste from mining

Conceptual design studies for the European DEMO divertor: Rationale and first results

Czech Academy of Sciences Publication Activity Database

You, J.H.; Mazzone, F.; Visca, E.; Bachmann, Ch.; Autissier, E.; Barrett, T.; Cocilovo, V.; Crescenzi, F.; Domalapally, P.K.; Dongiovanni, D.; Entler, Slavomír; Federici, G.; Frosi, P.; Fursdon, M.; Greuner, H.; Hancock, D.; Marzullo, D.; McIntosh, S.; Müller, A.V.; Porfiri, M.T.; Ramogida, G.; Reiser, J.; Richou, M.; Rieth, M.; Rydzy, A.; Villari, R.; Widak, V.

109-111, November (2016), s. 1598-1603 ISSN 0920-3796. [International Symposium on Fusion Nuclear Technology (ISFNT-12)/12./. Jeju, 14.09.2015-18.09.2015] EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : DEMO * Tokamak * Divertor * Plasma-facing component * Conceptual design * Eurofusiona Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379615303331

ITC18: 18th international Toki conference. Development of physics and technology of stellarators/heliotrons 'en route to DEMO'. Proceedings

International Nuclear Information System (INIS)

2009-02-01

18th International Toki Conference (ITC18) was held in Toki (Japan) December 9-12 2008 organized by the National Institute for Fusion Science (NIFS). More than 150 experts in fusion research, especially in stellarator/heliotron research from Australia, Belgium, China, France, Germany, Hungary, India, Iran, Italy, Japan, Korea, Serbia, Spain, Sweden, Switzerland, and the United States of America gathered at the conference. The International Organizing Committee (IOC) chaired by O. Motojima, the International Program Committee (IPC) chaired by Y. Ogawa and the Local Organizing Committee (LOC) chaired by T. Mutoh have played the leading role in the elaboration of the scientific program of the conference. NIFS has organized the ITC as an annual meeting for fusion related sciences since its establishment in 1989. The IPC arranged 2 plenary talks, 1 review talk, 34 invited talks in addition to 109 contributed presentations including 6 oral talks. Recent developments in the experimental, theoretical and technical research show the clear route to the realization of a stellarator/heliotron type demo fusion reactor. ITC18 was devoted to review the recent developments and to discuss the next steps forward to the demo reactor realization of stellarator/heliotron type. In the conference, recent experimental results from both tokamak and stellarator/heliotron devices are reviewed and the experimental and theoretical physics of plasma confinement in toroidal devices are also discussed and confirmed that the physical base of the fusion reactor is well developed. The development of steady state operation, heating, fueling, divertors, plasma wall interaction and wall materials, advanced diagnostics for reactor relevant plasma, blanket materials as well as super conducting magnets are discussed as inevitable key physics and technologies for the DEMO reactor. Slides of all oral presentations as well as the proceedings are available at http://itc.nifs.ac.jp/. Extended papers of major

Development status of the integrated tokamak simulator for K-DEMO

Energy Technology Data Exchange (ETDEWEB)

Kang, J. S.; Wang, J.; Hwang, Y. S. [Seoul National University, Seoul (Korea, Republic of); Jung, L. [National Fusion Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Korean fusion demonstration reactor (K-DEMO) study has been conducted to investigate the feasibility of an electricity generation, self-sustained tritium cycle, and component test facility. To estimate its capability, the integrated fusion operation simulator called INFRA has been developed by organizing relevant computational codes with standard data models and framework. The different modules of the integrated simulator are chosen among well-validated codes. Standard data models are directly linked with KSTAR experimental data so that the integrated simulator can be used for interpretative simulations but also for predictive simulations. In this study, the current status of code development and some examples of KSTAR interpretative simulations are reported. ITER integrated modelling and analysis suite is imported to K-DEMO data model to take over ITER experience and to accelerate collaboration with international IMAS community. Standardized rules and guideline have been developed by ITER team for many years. Based on strict policy, this data model has been established and updated. This data model is used for experimental and simulation results. The INFRA system has been utilized to be an alpha version of a KDEMO simulator. Database, framework, and module integration are conducted. A test equilibrium run for KSTAR is done by filling the database with experiment results. More modules will be incorporated in a near future. Validation with KSTAR data and benchmarking previous modelling activity is also planned in order to confirm the feasibility of this system.

Tokamak DEMO-FNS: Concept of magnet system and vacuum chamber

Energy Technology Data Exchange (ETDEWEB)

Azizov, E. A., E-mail: Azizov-EA@nrcki.ru; Ananyev, S. S. [National Research Center Kurchatov Institute (Russian Federation); Belyakov, V. A.; Bondarchuk, E. N.; Voronova, A. A. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Golikov, A. A. [National Research Center Kurchatov Institute (Russian Federation); Goncharov, P. R. [Peter the Great St. Petersburg Polytechnic University (Russian Federation); Dnestrovskij, A. Yu. [National Research Center Kurchatov Institute (Russian Federation); Zapretilina, E. R. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Ivanov, D. P. [National Research Center Kurchatov Institute (Russian Federation); Kavin, A. A.; Kedrov, I. V. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Klischenko, A. V.; Kolbasov, B. N. [National Research Center Kurchatov Institute (Russian Federation); Krasnov, S. V. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Krylov, A. I. [National Research Center Kurchatov Institute (Russian Federation); Krylov, V. A.; Kuzmin, E. G. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); Kuteev, B. V. [National Research Center Kurchatov Institute (Russian Federation); Labusov, A. N. [D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (Russian Federation); and others

2016-12-15

The level of knowledge accumulated to date in the physics and technologies of controlled thermonuclear fusion (CTF) makes it possible to begin designing fusion—fission hybrid systems that would involve a fusion neutron source (FNS) and which would admit employment for the production of fissile materials and for the transmutation of spent nuclear fuel. Modern Russian strategies for CTF development plan the construction to 2023 of tokamak-based demonstration hybrid FNS for implementing steady-state plasma burning, testing hybrid blankets, and evolving nuclear technologies. Work on designing the DEMO-FNS facility is still in its infancy. The Efremov Institute began designing its magnet system and vacuum chamber, while the Kurchatov Institute developed plasma-physics design aspects and determined basic parameters of the facility. The major radius of the plasma in the DEMO-FNS facility is R = 2.75 m, while its minor radius is a = 1 m; the plasma elongation is k{sub 95} = 2. The fusion power is P{sub FUS} = 40 MW. The toroidal magnetic field on the plasma-filament axis is B{sub t0} = 5 T. The plasma current is I{sub p} = 5 MA. The application of superconductors in the magnet system permits drastically reducing the power consumed by its magnets but requires arranging a thick radiation shield between the plasma and magnet system. The central solenoid, toroidal-field coils, and poloidal-field coils are manufactured from, respectively, Nb{sub 3}Sn, NbTi and Nb{sub 3}Sn, and NbTi. The vacuum chamber is a double-wall vessel. The space between the walls manufactured from 316L austenitic steel is filled with an iron—water radiation shield (70% of stainless steel and 30% of water).

The spherical tokamak fusion power plant

International Nuclear Information System (INIS)

Wilson, H.R.; Voss, G.; Ahn, J.W.

2003-01-01

The design of a 1GW(e) steady state fusion power plant, based on the spherical tokamak concept, has been further iterated towards a fully self-consistent solution taking account of plasma physics, engineering and neutronics constraints. In particular a plausible solution to exhaust handling is proposed and the steam cycle refined to further improve efficiency. The physics design takes full account of confinement, MHD stability and steady state current drive. It is proposed that such a design may offer a fusion power plant which is easy to maintain: an attractive feature for the power plants following ITER. (author)

On Korean strategy and plan for fusion energy

International Nuclear Information System (INIS)

Kim, H.J.; Choi, W-J.; Park, C.; Kim, H.C.

2012-01-01

In developing KSTAR (Korean Superconducting Tokamak Advanced Research), Korea had initiated a mid-entry strategy to catch up with the technologies required for the development of a fusion reactor, based on the tokamak magnetic confinement concept. Upon joining ITER (International Thermonuclear Experimental Reactor), Korean government enacted a promotional law for the fusion energy development. Under this promotional law the national promotional plans for developing fusion energy have been established. The National Fusion Research Institute (NFRI) developed the strategy and plan for a fusion DEMO program to realize the magnetic fusion energy. (author)

On Korean strategy and plan for fusion energy

Energy Technology Data Exchange (ETDEWEB)

Kim, H.J. [National Fusion Research Inst., Daejeon (Korea, Republic of); Choi, W-J. [Chungnam National Univ., Daejeon (Korea, Republic of); Park, C. [POSTECH, Pohang (Korea, Republic of); Kim, H.C. [National Fusion Research Inst., Daejeon (Korea, Republic of)

2012-07-01

In developing KSTAR (Korean Superconducting Tokamak Advanced Research), Korea had initiated a mid-entry strategy to catch up with the technologies required for the development of a fusion reactor, based on the tokamak magnetic confinement concept. Upon joining ITER (International Thermonuclear Experimental Reactor), Korean government enacted a promotional law for the fusion energy development. Under this promotional law the national promotional plans for developing fusion energy have been established. The National Fusion Research Institute (NFRI) developed the strategy and plan for a fusion DEMO program to realize the magnetic fusion energy. (author)

Physics, systems analysis and economics of fusion power plants

International Nuclear Information System (INIS)

Ward, D.J.

2006-01-01

Fusion power is being developed because of its large resource base, low environmental impact and high levels of intrinsic safety. It is important, however, to investigate the economics of a future fusion power plant to check that the electricity produced can, in fact, have a market. Using systems code analysis, including costing algorithms, this paper gives the cost of electricity expected from a range of fusion power plants, assuming that they are brought into successful operation. Although this paper does not purport to show that a first generation of fusion plants is likely to be the cheapest option for a future energy source, such plants look likely to have a market in some countries even without taking account of fusion's environmental advantages. With improved technological maturity fusion looks likely to have a widespread potential market particularly if the value of its environmental advantages are captured, for instance through avoiding a carbon tax. (author)

The Long way Towards Inertial Fusion Energy (lirpp Vol. 13)

Science.gov (United States)

Velarde, Guillermo

2016-10-01

In 1955 the first Geneva Conference was held in which two important events took place. Firstly, the announcement by President Eisenhower of the Program Atoms for Peace declassifying the information concerning nuclear fission reactors. Secondly, it was forecast that due to the research made on stellerators and magnetic mirrors, the first demo fusion facility would be in operation within ten years. This forecasting, as all of us know today, was a mistake. Forty years afterwards, we can say that probably the first Demo Reactor will be operative in some years more and I sincerely hope that it will be based on the inertial fusion concept...

Neutronics studies for the design of the European DEMO vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Flammini, Davide, E-mail: davide.flammini@enea.it [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Villari, Rosaria; Moro, Fabio; Pizzuto, Aldo [ENEA, Fusion Technical Unit, Nuclear Technologies Laboratory, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Bachmann, Christian [EUROfusion Consortium, Boltzmannstr. 2, 85748 Garching (Germany)

2016-11-01

Highlights: • MCNP calculation of nuclear heating, damage, helium production and neutron flux in DEMO HCLL and HCPB vacuum vessel at the inboard equatorial plane. • Study of impact of the poloidal gap between blanket modules, for several gap width, on vacuum vessel nuclear quantities. • Effect of the gap on nuclear heating result to be moderate, however high values of nuclear heating are found, even far from the gap with HCLL blanket. • Radiation damage limit of 2.75 DPA is met with a 1 cm wide gap. Helium production results very sensitive to the gap width. • Comparison between HCLL and HCPB blankets is shown for nuclear heating and neutron flux in the vacuum vessel. - Abstract: The DEMO vacuum vessel, a massive water cooled double-walled steel vessel, is located behind breeding blankets and manifolds and it will be subjected to an intense neutron and photon irradiation. Therefore, a proper evaluation of the vessel nuclear heat loads is required to assure adequate cooling and, given the significant lifetime neutron fluence of DEMO, the radiation damage limit of the vessel needs to be carefully controlled. In the present work nuclear heating, radiation damage (DPA), helium production, neutron and photon fluxes have been calculated on the vacuum vessel at the inboard by means of MCNP5 using a 3D Helium Cooled Lithium Lead (HCLL) DEMO model with 1572 MW of fusion power. In particular, the effect of the poloidal gap between the breeding-blanket segments on vacuum vessel nuclear loads has been estimated varying the gap width from 0 to 5 cm. High values of the nuclear heating (≈1 W/cm{sup 3}), which might cause intense thermal stresses, were obtained in inboard equatorial zone. The effect of the poloidal gap on the nuclear heating resulted to be moderate (within 30%). The radiation damage limit of 2.75 DPA on the vessel is almost met with 1 cm of poloidal gap over DEMO lifetime. A comparison with Helium Cooled Pebble Bed blanket is also provided.

Development of Tokamak reactor system code and conceptual studies of DEMO with He Cooled Molten Li blanket

International Nuclear Information System (INIS)

Hong, B.G.; Lee, Dong Won; Kim, Yong Hi

2007-01-01

To develop the concepts of fusion power plants and identify the design parameters, we have been developing the tokamak reactor system code. The system code can take into account a wide range of plasma physics and technology effects simultaneously and it can be used to find design parameters which optimize the given figure of merits. The outcome of the system studies using the system code is to identify which areas of plasma physics and technologies and to what extent should be developed for realization of a given fusion power plant concepts. As an application of the tokamak reactor system code, we investigate the performance of DEMO for early realization with a limited extension from the plasma physics and technology used in the design of the ITER. Main requirements for DEMO are selected as: 1) to demonstrate tritium self-sufficiency, 2) to generate net electricity, and 3) for steady-state operation. The size of plasma is assumed to be same as that of ITER and the plasma parameters which characterize the performance, i.e. normalized β value, β N , confinement improvement factor for the H-mode, H and the ratio of the Greenwald density limit n/n G are assumed to be improved beyond those of ITER: β N >2.0, H>1.0 and n/n G >1.0. Tritium self-sufficiency is provided by the He Cooled Molten Lithium (HCML) blanket with the total thickness of 2.5 m including the shield. With n/n G >1.2, net electric power bigger than 500 MW is possible with β N >4.0 andH>1.2. To access operation space for higher electric power, main restrictions are given by the divertor heat load and the steady-state operation requirements. Developments in both plasma physics and technology are required to handle high heat load and to increase the current drive efficiency. (orig.)

ITER: the Sun rises over nuclear fusion with West

International Nuclear Information System (INIS)

Sacco, Laurent

2013-01-01

The ITER project is considered as a critical step on the way to commercial production of electricity by a thermonuclear reactor based on controlled fusion. This project notably requires the development of a divertor which is the objective of the West project which will use the famous Cadarache superconductive magnet reactor, Tore Supra. After having outlined the future lack of fossil energies at the world scale, presented the operation principles of tokamaks and recalled some results obtained in their development, this article justifies the use of superconductive magnets. It presents the ITER project as a step in the production of thermonuclear electricity. ITER will be in fact a proof that such plants can be realised, and it should be followed by Demo, a demonstration power plant, by 2050. The article presents the West project, a test bench for ITER, which introduced modifications in the Tore Supra reactor to create conditions almost similar to that existing at the surface of the Sun. It notably comprises a divertor made of tungsten for the fusion with tritium. It finally outlines that the fusion will be a hot one, not a cold one

Energy, material and land requirement of a fusion plant

DEFF Research Database (Denmark)

Schleisner, Liselotte; Hamacher, T.; Cabal, H.

2001-01-01

The energy and material necessary to construct a power plant and the land covered by the plant are indicators for the ‘consumption’ of environment by a certain technology. Based on current knowledge, estimations show that the material necessary to construct a fusion plant will exceed the material...... requirement of a fission plant by a factor of two. The material requirement for a fusion plant is roughly 2000 t/MW and little less than 1000 t/MW for a fission plant. The land requirement for a fusion plant is roughly 300 m2/MW and the land requirement for a fission plant is a little less than 200 m2/MW...... less ‘environment’ for the construction than renewable technologies, especially wind and solar....

Advanced materials characterization and modeling using synchrotron, neutron, TEM, and novel micro-mechanical techniques - A European effort to accelerate fusion materials development

DEFF Research Database (Denmark)

Linsmeier, Ch.; Fu, C.-C.; Kaprolat, A.

2013-01-01

as testing under neutron flux-induced conditions. For the realization of a DEMO power plant, the materials solutions must be available in time. The European initiative FEMaS-CA – Fusion Energy Materials Science – Coordination Action – aims at accelerating materials development by integrating advanced...... having energies up to 14 MeV. In addition to withstanding the effects of neutrons, the mechanical stability of structural materials has to be maintained up to high temperatures. Plasma-exposed materials must be compatible with the fusion plasma, both with regard to the generation of impurities injected...

The plant efficiency of fusion power stations

International Nuclear Information System (INIS)

Darvas, J.; Foerster, S.

1976-01-01

Due to the circulating energy, lower efficiencies are to be expected with fusion power plants than with nuclear fission power plants. According to the systems analysis, the mirror machine is not very promising as a power plant. The plant efficiency of the laser fusion strongly depends on the laser efficiency about which one can only make speculative statements at present. The Tokamak requires a relatively low circulating energy and is certainly able to compete regarding efficiency as long as the consumption time can be kept large (> 100 sec) and the dead time between the power pulses small ( [de

Afterheat assessment of a conceptual fusion reactor

International Nuclear Information System (INIS)

Jayatissa, S.P.; Goddard, A.J.H.

1987-01-01

Structural activation and decay heat deposition calculations have been undertaken for the DEMO fusion reactor design. The DEMO design was based on an earlier conceptual design of a blanket sector which could breed tritium and generate electricity. These calculations have taken account of the redistribution of energy by the transport of γ radiation. Calculated heat deposition patterns have been used as data for simplified heat transfer calculations to judge temperature rises in relation to materials limits in a severe accident involving complete coolant flow failure. (author)

Fusion power plant availability study

International Nuclear Information System (INIS)

Ladra, D.; Sanguinetti, G.P.; Stube, E.

2001-01-01

The consideration of fusion as an alternative energy source will need to demonstrate that Fusion Power Plant (FPP) design, operating and maintenance characteristics meet the electrical market requirements forecast for the second half of this century. Until now, fusion has been developed in the framework of research and development programmes following natural technological trends. To bring a greater sense of realism to commercial viability and to guarantee that technology-driven fusion development responds to the demands of the market, a conceptual study of future commercial FPPs has been performed with a Power Plant Availability (PPA) study aimed at identifying the aspects affecting the availability and generating costs of FPPs. EFET, who has also been involved in the study, can visualise it from two different points of view; that of the industry (ANSALDO, IBERTEF, SIEMENS, NNC) and that of the utilities (BELGATOM, FRAMATOME, FORTUM). The work carried out covered the following points: socio-economic forecasting; safety and licensing; operation and maintenance; waste and decommissioning; availability and reliability. The following are the most relevant findings, conclusions and recommendations for all these aspects: Demonstrate definitively that the physical principles of nuclear fusion have been validated by means of experiments; Establish a European Industrial Group to support the demonstration phases; Create the financial and contracting framework required to construct these installations. Secure the necessary budgets for the European Union's 5th and 6th Research Programmes. Look for supplementary long term financing sources; The existing Regulatory Bodies should combine to form a single Working Group with responsibility for fusion reactor safety and licensing activities, working on the harmonisation of the regulatory processes, developing FPP safety criteria and guidelines and reviewing industry standards; To be competitive, FPPs should have high availability

Fusion Programme SCK-CEN - Annual Report 2010

Energy Technology Data Exchange (ETDEWEB)

Massaut, V.

2010-10-15

In 2010 SCK-CEN centred his activities in fusion Research and Development around four main poles: 1) the studies of the first wall of future DEMO facility and plasma wall interactions, using the recently refurbished plasmatron VISIONI; 2) the study of the radiation resistance of optical and specific diagnostic components and the development and prototyping of a Fiber Optics Current Sensor (FOCS) for measuring Tokamak plasma current for long plasma pulses and without embarked electronics: 3) the further study, by irradiation and mechanical testing but also by modelling, of the future structural material for a fusion plant, the RAFM Eurofer and the development and characterization of ODS-Eurofer; 4) specific fusion socio-economic studies on fusion, based on the specific developments carried out for fusion and radioactive waste management. SCK-CEN is also strongly involved in the Broader Approach agreement as Designated Institution for the Belgian State, having a coordinating and managing role for all Belgian activities in this agreement. The present report is structured following the work programme 2010 of the Association. Some former activities form the old-EFDA have been grouped with new ones in coherent and collaborative packages. Most activities of SCK-CEN are, and have always been, carried out under EFDA task agreements.

Fusion Programme SCK-CEN - Annual Report 2010

International Nuclear Information System (INIS)

Massaut, V.

2010-01-01

In 2010 SCK-CEN centred his activities in fusion Research and Development around four main poles: 1) the studies of the first wall of future DEMO facility and plasma wall interactions, using the recently refurbished plasmatron VISIONI; 2) the study of the radiation resistance of optical and specific diagnostic components and the development and prototyping of a Fiber Optics Current Sensor (FOCS) for measuring Tokamak plasma current for long plasma pulses and without embarked electronics: 3) the further study, by irradiation and mechanical testing but also by modelling, of the future structural material for a fusion plant, the RAFM Eurofer and the development and characterization of ODS-Eurofer; 4) specific fusion socio-economic studies on fusion, based on the specific developments carried out for fusion and radioactive waste management. SCK-CEN is also strongly involved in the Broader Approach agreement as Designated Institution for the Belgian State, having a coordinating and managing role for all Belgian activities in this agreement. The present report is structured following the work programme 2010 of the Association. Some former activities form the old-EFDA have been grouped with new ones in coherent and collaborative packages. Most activities of SCK-CEN are, and have always been, carried out under EFDA task agreements.

Complexity and availability for fusion power plants: The potential advantages of inertial fusion energy

International Nuclear Information System (INIS)

Perkins, L.J.

1997-01-01

Probably the single largest advantage of the inertial route to fusion energy (IFE) is the perception that its power plant embodiments could achieve acceptable capacity factors. This is a result of its relative simplicity, the decoupling of the driver and reactor chamber, and the potential to employ thick liquid walls. The author examines these issues in terms of the complexity, reliability, maintainability and, therefore, availability of both magnetic and inertial fusion power plants and compares these factors with corresponding scheduled and unscheduled outage data from present day fission experience. The author stresses that, given the simple nature of a fission core, the vast majority of unplanned outages in fission plants are due to failures outside the reactor vessel itself. Given one must be prepared for similar outages in the analogous plant external to a fusion power core, this puts severe demands on the reliability required of the fusion core itself. The author indicates that such requirements can probably be met for IFE plants. He recommends that this advantage be promoted by performing a quantitative reliability and availability study for a representative IFE power plant and suggests that databases are probably adequate for this task. 40 refs., 4 figs., 3 tabs

Energetic-economic analysis of inertial fusion plants with tritium commercial production

International Nuclear Information System (INIS)

Vezzani, M.; Cerullo, N.; Lanza, S.

2000-01-01

The realization of nuclear power plants based on fusion principles is expected to be, at the moment, very expensive. As a result the expected cost of electricity (COE) of fusion power plants is much higher than the COE of fission and fossil power plants. Thus it is necessary to study new solutions for fusion power plant designs to reduce the COE. An interesting solution for the first generation of fusion plants is to produce a surplus of tritium for commercial purposes. The present paper is concerned with the study of whether such a tritium surplus production can improve the plant economic balance, so that the COE is reduced, and to what extent. The result was that such a production allows a considerable reduction of COE and seems to be a good direction for development for the first generation of fusion power plants. To give an example, for a reference inertial confinement fusion (ICF) power plant the rise of the plant net tritium breeding ratio (TBR n ) from 1 to 1.2 would allow, in the conservative estimate of a tritium market price (C T ) of 5 M$/kg, a COE reduction of about 20%. In the estimate of a TBR n rise from 1 to 1.3 and of a C T value of 10 M$/kg, COE reduction could be more than 50%! In conclusion, the present paper points out the influence of TBR increase on COE reduction. Such a conclusion, which holds true for every fusion plant, is much more valid for ICF plants in which it is possible to reach higher TBR values and to use tritium extraction systems easily. Thus, considering the relevant economic advantages, a commercial tritium surplus production should not be disregarded for first generation fusion power plant designs, in particular for ICF plant designs

ECN's torrefaction-based BO2-technology. From pilot to demo

Energy Technology Data Exchange (ETDEWEB)

Kiel, J.H.A. [ECN Biomass, Coal and Environmental Research, Petten (Netherlands)

2011-02-15

The contents of this PowerPoint presentation are: Torrefaction design challenges; Initial small-scale R and D; ECN's torrefaction-based BO2-technology; Pilot-scale testing; and Demonstration and market introduction. The conclusions state that Torrefaction potentially allows cost-effective production of 2nd generation biomass pellets from a wide range of biomass/waste feedstock with a high energy efficiency (>90%); Torrefaction pellets show: High energy density, Water resistance, No/Limited biological degradation and heating, Excellent grindability, and Good combustion and gasification properties; Torrefaction is a separate thermal regime and requires dedicated reactor/process design; Torrefaction development is in pilot/demo-phase and shows strong market pull for torrefaction plants and torrefaction pellets; For ECN's BO2-technology a demo-plant is in preparation and industrial partnership for world-wide market introduction is nearly established.

Trends in radiation protection: possible effects on fusion power plant design

International Nuclear Information System (INIS)

Eurajoki, Tapani; Frias, Manuel Pascual; Orlandi, Sergio

2003-01-01

Since the design of fusion power plants involves long-term issues, ranging over several decades, it is useful to try to foresee under what kind of regulations the first fusion plants are to be operated. Application of present radiological regulations and practice to a fusion power plant concept is considered. The current design phase of fusion power plants motivates the top-down dose assessment, but it is crucial to aim at bottom-up assessments to ensure radiation doses as low as reasonably achievable. Since several issues, relating both to our knowledge on radiation as well as to the practice of radiation protection, may change in the future, it is necessary to continuously follow the development in the further design of fusion power plants

A Feasible DEMO Blanket Concept Based on Water Cooled Solid Breeder

Energy Technology Data Exchange (ETDEWEB)

Someya, Y.; Tobita, K.; Utoh, H.; Hoshino, K.; Asakura, N.; Nakamura, M.; Tanigawa, H.; Mikio, E.; Tanigawa, H.; Nakamichi, M.; Hoshino, T., E-mail: someya.yoji@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

2012-09-15

Full text: JAEA has conducted the conceptual design study of blanket for a fusion DEMO reactor SlimCS. Considering DEMO specific requirements, we place emphasis on a blanket concept with durability to severe irradiation, ease of fabrication for mass production, operation temperature of blanket materials, and maintainability using remote handling equipment. This paper present a promising concept satisfying these requirements, which is characterized by minimized welding lines near the front, a simplified blanket interior consisting of cooling tubes and a mixed pebble bed of breeder and neutron multiplier, and approximately the same outlet temperature for all blanket modules. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in tritium breeding ratio even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production, which will facilitate the access of remote handling equipment for replacement of the blanket modules and improve the access of diagnostics. (author)

Evaluation of the energy required for constructing and operating a fusion power plant

International Nuclear Information System (INIS)

Buende, R.

1982-09-01

The energy required for constructing and operating a tokamak fusion power plant is appraised with respect to the energy output during the lifetime of the plant. A harvesting factor is deduced as a relevant figure of energetic merit and is used for a comparison between fusion, fission, and coal-fired power plants. Because fusion power plants involve considerable uncertainties the comparison is supplemented by a sensitivity analysis. In comparison with Light Water Reactor plants fusion power plants appear to be rather favourable in this respect. The energy required for providing the fuel is relatively low for fusion plants, thus overcompensating the considerable higher amount of energy necessary for constructing the fusion power plant. (orig.)

Advanced fission and fossil plant economics-implications for fusion

International Nuclear Information System (INIS)

Delene, J.G.

1994-01-01

In order for fusion energy to be a viable option for electric power generation, it must either directly compete with future alternatives or serve as a reasonable backup if the alternatives become unacceptable. This paper discusses projected costs for the most likely competitors with fusion power for baseload electric capacity and what these costs imply for fusion economics. The competitors examined include advanced nuclear fission and advanced fossil-fired plants. The projected costs and their basis are discussed. The estimates for these technologies are compared with cost estimates for magnetic and inertial confinement fusion plants. The conclusion of the analysis is that fusion faces formidable economic competition. Although the cost level for fusion appears greater than that for fission or fossil, the costs are not so high as to preclude fusion's potential competitiveness

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

Multiple Module Simulation of Water Cooled Breeding Blankets in K-DEMO Using Thermal-Hydraulic Analysis Code MARS-KS

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon-Woo; Lee, Jeong-Hun; Park, Goon-Cherl; Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

2015-10-15

A preliminary concept for the Korean fusion demonstration reactor (K-DEMO) has been studied by the National Fusion Research Institute (NFRI) based on the National Fusion Roadmap of Korea. The feasibility studies have been performed in order to establish the conceptual design guidelines of the breeding blanket. As a part of the NFRI research, Seoul National University (SNU) is conducting thermal design, evaluation and validation of the water-cooled breeding blanket for the K-DEMO reactor. The purpose of this study is to extend the capability of MARS-KS to the overall blanket system analysis which includes 736 blanket modules in total. The strategy for the multi-module blanket system analysis using MARS-KS is introduced and the analysis result of the 46 blanket modules of single sector was summarized. A thermal-hydraulic analysis code for a nuclear reactor safety, MARS-KS, was applied for thermal analysis of the conceptual design of the K-DEMO breeding blanket. Then, a methodology to simulate multiple blanket modules was proposed, which uses a supervisor program to handle each blanket module individually at first and then distribute the flow rate considering the pressure drop that occurs in each module. For a feasibility test of the proposed methodology, 46 blankets in a sector, which are connected with each other through the common headers for the sector inlet and outlet, were simulated. The calculation results of flow rates, pressure drops, and temperatures showed the validity of the calculation. Because of parallelization using the MPI system, the computational time could be reduced significantly. In future, this methodology will be extended to an efficient simulation of multiple sectors, and further validation for transient simulation will be carried out for more practical applications.

On the definition of a DEMO (demonstration) reactor

International Nuclear Information System (INIS)

Cole, H.C.; Challender, R.S.

1987-01-01

The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. The authors have suggested a definition of a DEMO, and listed what they considered to be the most important implications of this definition. A table of parameters is included comparing published DEMO's with typical commercial reactor and 'pre-DEMO' studies. (U.K.)

The path from ITER to a power plant - initial results from the ARIES ''Pathways'' program

International Nuclear Information System (INIS)

Najmabadi, F.

2007-01-01

The US national power plant studies program, ARIES, has initiated a 3-year integrated study, called the ''Pathways Program'' to investigate what the fusion program needs to do, in addition to successful operation of the ITER, in order to transform fusion into a commercial reality. The US power industry and regulatory agencies view the demonstration power plant, DEMO, as a device which is build and operated by industry, possibly with government participation, to demonstrate the commercial readiness of fusion power. As such, the ''Pathways'' programs will investigate what is needed, in addition to successful operation of ITER, to convince industry to move forward with a fusion DEMO. While many reports exists that provide a strategic view of the needs for fusion development; in the ITER era, a much more detailed view is needed to provide the necessary information for program planning. By comparing the anticipated results from ITER and existing facilities with the requirements for a power plant in the first phase of the Pathways study, we will develop a comprehensive list of remaining R and D items for developing fusion, will identify metrics for distributing resources among R and D issues, and will identify which of those items can/should be done in existing or simulation facilities. In the second phase of the study, we will develop potential embodiments for the fusion test facility (ies) and explore their cost/performance parametrically. An important by-product of this study is the identification of key R and D issues that can be performed and resolved in existing facilities to make the fusion facility cheaper and/or a higher performance device. This paper summarizes the results from the first phase of our study. We have adopted a ''holistic'' or integrated approach with the focus on the needs of the customer. In such an approach, the remaining R and D should generate all of the information needed by industry to move forward with the DEMO, i.e., data needed to

Structural integrity for DEMO: An opportunity to close the gap from materials science to engineering needs

Energy Technology Data Exchange (ETDEWEB)

Porton, M., E-mail: michael.porton@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Wynne, B.P. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); University of Sheffield, Sheffield, South Yorkshire S10 2TN (United Kingdom); Bamber, R.; Hardie, C.D.; Kalsey, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

2016-11-01

Highlights: • Key shortfalls in the current approaches to verification of structural integrity are outlined. • Case studies for high integrity applications in other demanding environments are examined. • Relevant lessons are drawn from fission and space for the design stage and through service life. • Future efforts are suggested to align materials and engineering for DEMO structural integrity. - Abstract: It is clear that fusion demonstration devices offer unique challenges due to the myriad, interacting material degradation effects and the numerous, conflicting requirements that must be addressed in order for in-vessel components to deliver satisfactory performance over the required lifetime. The link between mechanical engineering and materials science is pivotal to assure the timely realisation and exploitation of successful fusion power. A key aspect of this link is the verification of structural integrity, achieved at the design stage via structural design criteria against which designs are judged to be sufficiently resilient (or not) to failure, for a given set of loading conditions and desired lifetime. As various demonstration power plant designs progress through their current conceptual design phases, this paper seeks to highlight key shortfalls in this vital link between engineering needs and materials science, offering a perspective on where future attention can be prioritised to maximise impact. Firstly, issues in applying existing structural design criteria to demonstration power plant designs are identified. Whilst fusion offers particular challenges, there are significant insights to be gained from attempts to address such issues for high performance, high integrity applications in other demanding environments. Therefore case studies from beyond fusion are discussed. These offer examples where similar shortfalls have been successfully addressed, via approaches at the design stage and through service lifetime in order to deliver significant

Computer aided design of operational units for tritium recovery from Li17Pb83 blanket of a DEMO fusion reactor

International Nuclear Information System (INIS)

Malara, C.; Viola, A.

1995-01-01

The problem of tritium recovery from Li 17 Pb 83 blanket of a DEMO fusion reactor is analyzed with the objective of limiting tritium permeation into the cooling water to acceptable levels. To this aim, a mathematical model describing the tritium behavior in blanket/recovery unit circuit has been formulated. By solving the model equations, tritium permeation rate into the cooling water and tritium inventory in the blanket are evaluated as a function of dimensionless parameters describing the combined effects of overall resistance for tritium transfer from Li 17 Pb 83 alloy to cooling water, circulating rate of the molten alloy in blanket/recovery unit circuit and extraction efficiency of tritium recovery unit. The extraction efficiency is, in turn, evaluated as a function of the operating conditions of recovery unit. The design of tritium recovery unit is then optimized on the basis of the above parametric analysis and the results are herein reported and discussed for a tritium permeation limit of 10 g/day into the cooling water. 14 refs., 9 figs., 2 tabs

Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

International Nuclear Information System (INIS)

Neu, R.; Riesch, J.; Coenen, J.W.; Brinkmann, J.; Calvo, A.; Elgeti, S.; García-Rosales, C.; Greuner, H.; Hoeschen, T.; Holzner, G.; Klein, F.; Koch, F.

2016-01-01

Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W_f/W) has been developed incorporating extrinsic toughening mechanisms. Small W_f/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO_3 compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

Advanced tungsten materials for plasma-facing components of DEMO and fusion power plants

Energy Technology Data Exchange (ETDEWEB)

Neu, R., E-mail: Rudolf.Neu@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Fakultät für Maschinenbau, Technische Universität München, D-85748 Garching (Germany); Riesch, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Coenen, J.W. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Brinkmann, J. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Calvo, A. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Elgeti, S. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); García-Rosales, C. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Greuner, H.; Hoeschen, T.; Holzner, G. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); Klein, F. [Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik, D-52425 Jülich (Germany); Koch, F. [Max-Planck-Institut für Plasmaphysik, D-85748 Garching (Germany); and others

2016-11-01

Highlights: • Development of W-fibre enhanced W-composites incorporating extrinsic toughening mechanisms. • Production of a large sample (more than 2000 long fibres) for mechanical and thermal testing. • Even in a fully embrittled state, toughening mechanisms are still effective. • Emissions of volatile W-oxides can be suppressed by alloying W with elements forming stable oxides. • WCr10Ti2 has been successfully tested under accidental conditions and high heat fluxes. - Abstract: Tungsten is the major candidate material for the armour of plasma facing components in future fusion devices. To overcome the intrinsic brittleness of tungsten, which strongly limits its operational window, a W-fibre enhanced W-composite material (W{sub f}/W) has been developed incorporating extrinsic toughening mechanisms. Small W{sub f}/W samples show a large increase in toughness. Recently, a large sample (50 mm × 50 mm × 3 mm) with more than 2000 long fibres has been successfully produced allowing further mechanical and thermal testing. It could be shown that even in a fully embrittled state, toughening mechanisms as crack bridging by intact fibres, as well as the energy dissipation by fibre-matrix interface debonding and crack deflection are still effective. A potential problem with the use of pure W in a fusion reactor is the formation of radioactive and highly volatile WO{sub 3} compounds and their potential release under accidental conditions. It has been shown that the oxidation of W can be strongly suppressed by alloying with elements forming stable oxides. WCr10Ti2 alloy has been produced on a technical scale and has been successfully tested in the high heat flux test facility GLADIS. Recently, W-Cr-Y alloys have been produced on a lab-scale. They seem to have even improved properties compared to the previously investigated W alloys.

Evaluation of European blanket concepts for DEMO from availability and reliability point of view

International Nuclear Information System (INIS)

Nardi, C.

1995-12-01

This technical report is concerned with the ENEA activities relating to reliability and availability for the selection among two of the four European blanket concepts for the DEMO reactor. The activities on the BIT concept, the one proposed by ENEA, are emphasized. In spite of the lack of data relating to the behaviour of structures in an environment similar to that of a fusion reactor, it is evidenced that the available data are relevant to the BIT concept geometry. Moreover, it is evidenced that the qualitative reliability evaluations, compared to the quantitative ones, can lead to a better understanding of the typical problems of a structure to be used in a fusion reactor

Versatile Desktop Experiment Module (DEMo) on Heat Transfer

Science.gov (United States)

Minerick, Adrienne R.

2010-01-01

This paper outlines a new Desktop Experiment Module (DEMo) engineered for a chemical engineering junior-level Heat Transfer course. This new DEMo learning tool is versatile, fairly inexpensive, and portable such that it can be positioned on student desks throughout a classroom. The DEMo system can illustrate conduction of various materials,…

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

Science.gov (United States)

Fiorucci, Donatella; Hreibi, Ali; Chaibi, Walid

2018-03-01

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

Initial DEMO tokamak design configuration studies

Energy Technology Data Exchange (ETDEWEB)

Bachmann, Christian, E-mail: christian.bachmann@efda.org [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Aiello, G. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Albanese, R.; Ambrosino, R. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Arbeiter, F. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Aubert, J. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Boccaccini, L.; Carloni, D. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Federici, G. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kovari, M. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Li Puma, A. [CEA-Saclay, DEN, DM2S, SEMT, F-91191 Gif-Sur-Yvette (France); Loving, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Maione, I. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Mattei, M. [ENEA/CREATE, Universita di Napoli Federico II, Naples (Italy); Mazzone, G. [ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Meszaros, B. [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Palermo, I. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid (Spain); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Riccardo, V. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); and others

2015-10-15

Highlights: • A definition of main DEMO requirements. • A description of the DEMO tokamak design configuration. • A description of issues yet to be solved. - Abstract: To prepare the DEMO conceptual design phase a number of physics and engineering assessments were carried out in recent years in the frame of EFDA concluding in an initial design configuration of a DEMO tokamak. This paper gives an insight into the identified engineering requirements and constraints and describes their impact on the selection of the technologies and design principles of the main tokamak components. The EU DEMO program aims at making best use of the technologies developed for ITER (e.g., magnets, vessel, cryostat, and to some degree also the divertor). However, other systems in particular the breeding blanket require design solutions and advanced technologies that will only partially be tested in ITER. The main differences from ITER include the requirement to breed, to extract, to process and to recycle the tritium needed for plasma operation, the two orders of magnitude larger lifetime neutron fluence, the consequent radiation dose levels, which limit remote maintenance options, and the requirement to use low-activation steel for in-vessel components that also must operate at high temperature for efficient energy conversion.

Assessment of DEMO challenges in technology and physics

Energy Technology Data Exchange (ETDEWEB)

Zohm, Hartmut, E-mail: zohm@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany)

2013-10-15

Highlights: ► It is very important to respect the interlinks between physics and technology when developing designs for DEMO. ► Pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. ► Optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system is important. ► Exhaust requirements lead to an unprecedented high level of core radiation loss by impurity seeding in DEMO. -- Abstract: The challenges that DEMO designs encounter in both technology and physics are reviewed. It is shown that it is very important to respect the interlinks between these fields when developing designs for DEMO. Examples for areas where such interlinks put very strict requirements are the development of a steady state tokamak operation scenario and the question of power exhaust taking into account the boundary conditions set by materials questions. Concerning steady state operation, we find that demands on the physics scenario are so high that pulsed operation of a tokamak DEMO should seriously be considered in conservative DEMO designs. Alternatively, the device could foresee a large fraction of externally driven current which calls for optimization of both plasma CD efficiency as well as wall plug efficiency of the CD system. In the exhaust area, a realistic estimate of the admissable time averaged peak heat flux at the target is of the order of 5 MW/m{sup 2}, leading to strict requirements for the operational scenario, which has to rely on an unprecedented high level of radiation loss by impurity seeding and the facilitation of partial detachment. Thus, exhaust scenarios along these lines have to be developed which are compatible with the confinement needs and the H-L back transition power for DEMO. In both areas, we discuss possible risk mitigation strategies based on conceptually different approaches.

Some not such wonderful magnetic fusion facts; and their solution

Science.gov (United States)

Manheimer, Wallace

2017-10-01

The first not such wonderful fusion fact (NSWFF) is that if ITER is successful, it is nowhere near ready to develop into a DEMO. The design Q=10, along with electricity generating efficiency of 1/3 prevents this. Making it smaller and cheaper, increasing the gain by 3 or 4, and the wall loading by an order of magnitude is not a minor detail, it is not at all clear the success with ITER will lead to a similar, pure fusion DEMO. The second NSWFF is that tokamaks are unlikely to improve to the point where they can be effective fusion reactors because their performance is limited by conservative design rules. The third NSWFF is that developing large fusion devices like ITER takes an enormous amount of time and dollars, there are no second chances. The fourth NSWFF is that it is unlikely that alternative confinement configurations will succeed either, at least in this century; they are simply too far behind. There is only a single solution for fusion to become a sustainable, carbon free power source by midcentury or shortly thereafter. This is to develop ITER (assuming it is successful) into a fusion breeder. This work was not supported by any organization, private or public.

Fusion Programme SCK-CEN - Annual report 2009

International Nuclear Information System (INIS)

Massaut, V.

2009-01-01

This report summarizes the Research and Development work carried out at SCK-CEN on fusion technology in the year 2009. This covers mostly the work done under the EFDA agreement as well as the new developments carried out within the so-called Broader Approach of fusion such as - studies on structural and first wall materials for ITER and DEMO - studies and testing on the radiation resistance of instruments and componenets for the diagnostic and remote handling - development of irradiation devices and systems for the testing of fusion materials under representative environment.

Fusion Programme SCK-CEN - Annual report 2009

Energy Technology Data Exchange (ETDEWEB)

Massaut, V

2009-10-15

This report summarizes the Research and Development work carried out at SCK-CEN on fusion technology in the year 2009. This covers mostly the work done under the EFDA agreement as well as the new developments carried out within the so-called Broader Approach of fusion such as - studies on structural and first wall materials for ITER and DEMO - studies and testing on the radiation resistance of instruments and componenets for the diagnostic and remote handling - development of irradiation devices and systems for the testing of fusion materials under representative environment.

Implications of fusion power plant studies for materials requirements

International Nuclear Information System (INIS)

Cook, Ian; Ward, David; Dudarev, Sergei

2002-01-01

This paper addresses the key requirements for fusion materials, as these have emerged from studies of commercial fusion power plants. The objective of the international fusion programme is the creation of power stations that will have very attractive safety and environmental features and viable economics. Fusion power plant studies have shown that these objectives may be achieved without requiring extreme advances in materials. But it is required that existing candidate materials perform at least as well as envisaged in the environment of fusion neutrons, heat fluxes and particle fluxes. The development of advanced materials would bring further benefits. The work required entails the investigation of many intellectually exciting physics issues of great scientific interest, and of wider application than fusion. In addition to giving an overview, selected aspects of the science, of particular physics interest, are illustrated

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

On the NBI system for substantial current drive in a fusion power plant: Status and R and D needs for ion source and laser neutralizer

International Nuclear Information System (INIS)

Franzen, P.; Fantz, U.

2014-01-01

Highlights: • NBI is a candidate for a cw tokamak DEMO due to its high current drive efficiency. • The plug-in efficiency must be improved from the present 20–30% to more than 50%. • A suitable candidate is a photo neutralizer with almost 100% neutralization efficiency; basic feasibility studies are underway. • Cw operation with a large availability puts rather high demands on source operation with some safety margins, especially for the components with high power density loads (source back plate and extraction system). • Alternatives to the present use of cesium are under exploitations. - Abstract: The requirements for the heating and current drive systems of a fusion power plant will strongly depend on the DEMO scenario. The paper discusses the R and D needs for a neutral beam injection system — being a candidate due to the highest current drive efficiency — for the most demanding scenario, a steady state tokamak DEMO. Most important issues are the improvement of the wall-plug efficiency from the present ∼25% to the required 50–60% by improving the neutralization efficiency with a laser neutralizer system and the improvement of the reliability of the ion source operation. The demands on and the potential of decreasing the ion source operation pressure, as well as decreasing the amount of co-extracted electrons and backstreaming ions are discussed using the ITER requirements and solutions as basis. A further concern is the necessity of cesium for which either the cesium management must be improved or alternatives to cesium for the production of negative ions have to be identified

Superconductivity and fusion energy—the inseparable companions

Science.gov (United States)

Bruzzone, Pierluigi

2015-02-01

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

The European fusion program and the role of the research reactors

International Nuclear Information System (INIS)

Laesser, R.; Andreani, R.; Diegele, E.

2005-01-01

The main objectives of the European long-term Fusion Technology Program are i) investigation of DEMO breeding blankets options, ii) development of low activation materials resistant to high neutron fluence, iii) construction of IFMIF for validation of DEMO materials, and iv) promotion of modelling efforts for the understanding of radiation damage. A large effort is required for the development and performance verification of the materials subjected to the intense neutron irradiation encountered in fusion reactors. In the absence of a strong 14.1 MeV neutron source fission materials research reactors are used. Elaborate in-pile and post-irradiation examinations are performed. In addition, the modelling effort is increased to predict the damage by a 'true' fusion spectrum in the future. Even assuming that a positive decision for IFMIF construction can be reached, the operation of a limited number of materials test reactors is needed to perform irradiation studies on large samples and for screening. (author)

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

CERN Document Server

Daum, E

2000-01-01

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

Comparison study on neutronic analysis of the K-DEMO water cooled ceramic breeder blanket using MCNP and ATTILA

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Sung, E-mail: jspark@nfri.re.kr; Kwon, Sungjin; Im, Kihak

2016-11-01

Highlights: • A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA. • The calculation results of this study indicates that ATTILA showed close agreement with MCNP within ranges (3.3–28%). • Partly high discrepancy (17–28%) results between two codes existed to the nuclear heating calculation in high attenuating materials and radially thick structure regions. • The rest of the results showed small differences of NWL calculation (3.3%) and TBR distribution (3.9%). • ATTILA could be acceptable for K-DEMO neutronic analysis considering discrepancy (3.3–28%). - Abstract: A comparison study of main parameter calculations: neutron wall loading (NWL), tritium breeding ratio (TBR), and nuclear heating, on a Korean fusion demonstration reactor (K-DEMO) neutronic analysis model using MCNP and ATTILA was performed to investigate the feasibility of using ATTILA for the main parameter calculations. The model was created by commercial CAD program (Pro-Engineer™) as a 22.5° sector of tokamak consisting of major components such as blankets, shields, divertors, vacuum vessels (VV), toroidal field (TF) coils, and others, which was directly imported into ATTILA by Parasolid file. The discretizing in space, angle, and energy variables were refined for application of the K-DEMO neutronic analysis model through an iterative process since these variables greatly impact on accuracy, solution times, and memory consumptions in ATTILA. The main parameter calculations using ATTILA and the result of comparison studies indicate that the NWL distributions by two codes were almost agreed within discrepancy of 3.3%; the TBR distribution using ATTILA was slightly bigger than MCNP with a difference 3.9%; the nuclear heating values on TF coils and VV

Fusion reactor critical issues

International Nuclear Information System (INIS)

1987-11-01

The document summarizes the results of a series of INTOR-related meetings organized by the IAEA in 1985-1986 with the following topics: Impurity control modelling, non-inductive current-drive, confinement in tokamaks with intense heating and DEMO requirements. These results are useful to the specialists involved in research on large fusion machines or in the design activity on the next generation tokamaks. Refs, figs and tabs

Factors affecting potential market penetration of laser fusion power plants

International Nuclear Information System (INIS)

Deonigi, D.E.; Fraley, D.W.

1979-08-01

A mini-model has been constructed to estimate the optimal size of laser fusion power plants and to estimate the allowable cost of the first such plant in relation to the next best alternative. In estimating the costs of laser fusion, the mini-model incorporates such factors as market penetration, learning, economies of scale, system size, transmission costs, reserve requirements, development and licensing costs and site costs. The results of the mini-model simulations indicate that the optimal laser fusion plant size is approximately 3 GWe; risk considerations unincorporated in the mini-model suggest an optimal size closer to 2.5 GWe

Magnetic Fusion Energy Program of India

International Nuclear Information System (INIS)

Sen, Abhijit

2013-01-01

The magnetic fusion energy program of India started in the early eighties with the construction of an indigenous tokamak device ADITYA at the Institute for Plasma Research in Gandhinagar. The initial thrust was on fundamental studies related to plasma instabilities and turbulence phenomena but there was also a significant emphasis on technology development in the areas of magnetics, high vacuum, radio-frequency heating and neutral beam technology. The program took a major leap forward in the late nineties with the decision to build a state-of-the-art superconducting tokamak (SST-1) that catapulted India into the mainstream of the international tokamak research effort. The SST experience and the associated technological and human resource development has now earned the country a place in the ITER collaboration as an equal partner with other major nations. Keeping in mind the rapidly growing and enormous energy needs of the future the program has also identified and launched key development projects that can lead us to a DEMO reactor and eventually a Fusion Power Plant in a systematic manner. I will give a brief overview of the early origins, the present status and some of the highlights of the future road map of the Indian Fusion Program. (author)

A preliminary examination of the economics of cogeneration with fusion plants

International Nuclear Information System (INIS)

Hazelrigg, G.A.; Coleman, D.E.

1983-01-01

Cogeneration, the process of using reject heat from electric energy generation plants, offers substantial savings in energy consumption and thus is likely to see increased implementation, especially in the form of district heating, over the next few decades. The use of fusion plants for cogeneration offers added advantages of potentially low marginal costs and reduced siting restrictions compared to nuclear and coal plants, and freedom from use of limited fossil fuels. Fusion can thus provide increased economic incentive to the implementation of cogeneration systems. Conversely, cogeneration improves the economics of fusion and thus provides both added incentive for its development and reduced economic requirements on commercial fusion technologies

Materials-related issues in the safety and licensing of nuclear fusion facilities

Science.gov (United States)

Taylor, N.; Merrill, B.; Cadwallader, L.; Di Pace, L.; El-Guebaly, L.; Humrickhouse, P.; Panayotov, D.; Pinna, T.; Porfiri, M.-T.; Reyes, S.; Shimada, M.; Willms, S.

2017-09-01

Fusion power holds the promise of electricity production with a high degree of safety and low environmental impact. Favourable characteristics of fusion as an energy source provide the potential for this very good safety and environmental performance. But to fully realize the potential, attention must be paid in the design of a demonstration fusion power plant (DEMO) or a commercial power plant to minimize the radiological hazards. These hazards arise principally from the inventory of tritium and from materials that become activated by neutrons from the plasma. The confinement of these radioactive substances, and prevention of radiation exposure, are the primary goals of the safety approach for fusion, in order to minimize the potential for harm to personnel, the public, and the environment. The safety functions that are implemented in the design to achieve these goals are dependent on the performance of a range of materials. Degradation of the properties of materials can lead to challenges to key safety functions such as confinement. In this paper the principal types of material that have some role in safety are recalled. These either represent a potential source of hazard or contribute to the amelioration of hazards; in each case the related issues are reviewed. The resolution of these issues lead, in some instances, to requirements on materials specifications or to limits on their performance.

22nd IAEA Fusion Energy Conference: summary of sessions EX/D, EX/S and EX/W

International Nuclear Information System (INIS)

Motojima, O.

2009-01-01

The sessions of experiments EX/D, EX/S and EX/W covering (D) Plasma-material interactions, Divertors, Limiters, SOL, (S) Stability, (W) Wave-plasma interactions, Current drive, Heating and Energetic particles are summarized in this paper. These general topics are divided into two categories, which are (1) ITER oriented and (2) ITER/DEMO oriented corresponding to the subject of each topic and considering the road map of fusion research. Topics in the ITER oriented category are strongly linked to the present direction of the fusion research for ITER, namely tokamak research, whereas issues in the ITER/DEMO oriented category are generally common issues for both tokamaks and helical devices.

Development of liquid lithium divertor for fusion reactor

International Nuclear Information System (INIS)

Evtihkin, V. A.; Lyublinskij, I. E.; Vertkov, A.V.; Chumanov, A.V.; Shpolyanskij, V.N.

2000-01-01

Development of divertor is one of the most acute problems of the tokamak fusion reactor. The use of such materials as tungsten, beryllium, graphite and CFC's enabled to solve the problem to a certain extent fulfilling the need of the ITER project. The problem still rests unsolved for the DEMO-type reactors. Lithium if used as a material for high heat flux components may provide a successful solution of the problem. A concept of Li divertor based on the use of capillary-pore structures (CPS) is proposed and is being validated by a complex of experimental research and engineering developments. An optional concept of Li divertor for power removal at 400 MW in steady-state (DEMO-S project) is presented. The complex of experimental research is under way to prove the serviceability of the Li CPS in different conditions that would be realized in divertor

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)

The long range DIII-D plan

International Nuclear Information System (INIS)

Simonen, T.C.

1994-01-01

The mission of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. The National Energy Strategy calls for the development of magnetic fusion as an energy options with operation of a DEMO by 2025. The DEMO will be based on nuclear technology demonstrated in ITER and the physics and engineering database established in magnetic fusion facilities during the next two decades. On the present path, based on extrapolation of current conventional operating modes, ITER is twice as large as Joint European Tokamak (JET), and DEMO, using the same logic, will be even larger. However, successful development of advanced tokamak (AT) operating modes could open the way for significantly improved confinement and stability, leading to a smaller, more commercially attractive DEMO, provided new divertor concepts are developed to handle the accompanying high divertor power density. A smaller and lower cost DEMO opens up the possibility that multiple nations, utilities, and industries could build DEMOs simultaneously and, therefore, more rapidly optimize the tokamak for commercialization

Nuclear data for fusion technology – the European approach

Directory of Open Access Journals (Sweden)

Fischer Ulrich

2017-01-01

Full Text Available The European approach for the development of nuclear data for fusion technology applications is presented. Related R&D activities are conducted by the Consortium on Nuclear Data Development and Analysis for Fusion to satisfy the nuclear data needs of the major projects including ITER, the Early Neutron Source (ENS and DEMO. Recent achievements are presented in the area of nuclear data evaluations, benchmarking and validation, nuclear model improvements, and uncertainty assessments.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-10-15

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

Fusion reactor blanket-main design aspects

International Nuclear Information System (INIS)

Strebkov, Yu.; Sidorov, A.; Danilov, I.

1994-01-01

The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm 2 and 7 MWa/m 2 accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Effect of activation cross-section uncertainties on the radiological assessment of the MFE/DEMO first wall

Energy Technology Data Exchange (ETDEWEB)

Cabellos, O. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, Madrid (Spain)]. E-mail: cabellos@din.upm.es; Reyes, S. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Sanz, J. [Instituto de Fusion Nuclear, Universidad Politecnica de Madrid, Madrid (Spain); University Nacional Educacion a Distancia, Dep. Ingenieria Energetica, Juan del Rosal 12, 28040 Madrid (Spain); Rodriguez, A. [University Nacional Educacion a Distancia, Dep. Ingenieria Energetica, Juan del Rosal 12, 28040 Madrid (Spain); Youssef, M. [University of California, Los Angeles, CA (United States); Sawan, M. [University of Wisconsin, Madison, WI (United States)

2006-02-15

A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross-sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, {sup 6}Co and {sup 94}Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed.

Effect of activation cross-section uncertainties on the radiological assessment of the MFE/DEMO first wall

International Nuclear Information System (INIS)

Cabellos, O.; Reyes, S.; Sanz, J.; Rodriguez, A.; Youssef, M.; Sawan, M.

2006-01-01

A Monte Carlo procedure has been applied in this work in order to address the impact of activation cross-sections (XS) uncertainties on contact dose rate and decay heat calculations for the outboard first wall (FW) of a magnetic fusion energy (MFE) demonstration (DEMO) reactor. The XSs inducing the major uncertainty in the prediction of activation related quantities have been identified. Results have shown that for times corresponding to maintenance activities the uncertainties effect is insignificant since the dominant XSs involved in these calculations are based on accurate experimental data evaluations. However, for times corresponding to waste management/recycling activities, the errors induced by the XSs uncertainties, which in this case are evaluated using systematic models, must be considered. It has been found that two particular isotopes, 6 Co and 94 Nb, are key contributors to the global DEMO FW activation uncertainty results. In these cases, the benefit from further improvements in the accuracy of the critical reaction XSs is discussed

Progress in Developing a High-Availability Advanced Tokamak Pilot Plant

Energy Technology Data Exchange (ETDEWEB)

Brown, T.; Goldston, R.; Kessel, C.; Neilson, G.; Menard, J.; Prager, S.; Scott, S.; Titus, P.; Zarnstorff, M., E-mail: tbrown@pppl.gov [Princeton University, Princeton Plasma Physics Laboratory, Princeton (United States); Costley, A. [Henley on Thames (United Kingdom); El-Guebaly, L. [University of Wisconsin, Madison (United States); Malang, S. [Fusion Nuclear Technology Consulting, Linkenheim (Germany); Waganer, L. [St. Louis (United States)

2012-09-15

Full text: A fusion pilot plant study was initiated to clarify the development needs in moving from ITER to a first of a kind fusion power plant, following a path similar to the approach adopted for the commercialization of fission. The mission of the pilot plant was set to encompass component test and fusion nuclear science missions yet produce net electricity with high availability in a device designed to be prototypical of the commercial device. The objective of the study was to evaluate three different magnetic configuration options, the advanced tokamak (AT), spherical tokamak (ST) and compact stellarator (CS) in an effort to establish component characteristics, maintenance features and the general arrangement of each candidate device. With the move to look beyond ITER the fusion community is now beginning to embark on DEMO reactor studies with an emphasis on defining configuration arrangements that can meet a high availability goal. In this paper the AT pilot plant design will be presented. The selected maintenance approach, the device arrangement and sizing of the in-vessel components and details of interfacing auxiliary systems and services that impact the ability to achieve high availability operations will be discussed. Efforts made to enhance the interaction of in-vessel maintenance activities, the hot cell and the transfer process to develop simplifying solutions will also be addressed. (author)

Critical Design Factors for Sector Transport Maintenance in DEMO

Energy Technology Data Exchange (ETDEWEB)

Utoh, H.; Someya, Y.; Tobita, K.; Asakura, N.; Hoshino, K.; Nakamura, M., E-mail: uto.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan)

2012-09-15

Full text: Maintenance is a critical issue for fusion DEMO reactor because the design conditions and requirements of DEMO maintenance scheme are different from that of ITER remote handling. The sector transport maintenance scheme has advantages to maintain blankets and divertors without the use of sophisticated remote handling devices including sensitive devices to radiation in the reactor. SlimCS designed in JAEA adopts the sector transport maintenance scheme in which every sector is pulled out horizontally through a port between TF coils. A critical design issue for the horizontal sector transport maintenance scheme is how to support an enormous turnover force of the toroidal field (TF) coils. We propose following two options; first option is the horizontal transport maintenance scheme in which every sector is pulled out through four horizontal ports connected with the corridor. Second option is the vertical sector transport maintenance scheme with small vertical maintenance ports (total: 6 ports). The new horizontal sector transport limited in the number of maintenance ports is a more realistic maintenance scheme, and the key engineering issue is the transferring mechanism of sector in the vacuum vessel. In the maintenance scenario, the key design factors are the cool down time in reactor and the cooling method in maintenance scheme for keeping components under operation temperature. By one-dimensional heat conduction analysis, the sector should be transported to hot cell within 40 hours in the case the cool down time is one month. In the horizontal sector transport maintenance, the maintenance time including removal of cooling piping, drain of cooling water and sector transport to hot cell is about 32 hours. Furthermore, the tritium release in the sector transport can be suppressed because the components temperature drops by forced-air cooling system. This paper mainly focuses on a sector transport maintenance scheme from the aspects of high plant availability

Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

Science.gov (United States)

Vdovin, V.

2014-02-01

The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20-40) IC frequency harmonics) at frequencies of 500-1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure βN > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D - Kurchatov Institute experiment on helicons CD [1].

Current generation by helicons and LH waves in modern tokamaks and reactors FNSF-AT, ITER and DEMO. Scenarios, modeling and antennae

Energy Technology Data Exchange (ETDEWEB)

Vdovin, V. [NRC Kurchatov Institute Tokamak Physics Institute, Moscow (Russian Federation)

2014-02-12

The Innovative concept and 3D full wave code modeling Off-axis current drive by RF waves in large scale tokamaks, reactors FNSF-AT, ITER and DEMO for steady state operation with high efficiency was proposed [1] to overcome problems well known for LH method [2]. The scheme uses the helicons radiation (fast magnetosonic waves at high (20–40) IC frequency harmonics) at frequencies of 500–1000 MHz, propagating in the outer regions of the plasmas with a rotational transform. It is expected that the current generated by Helicons will help to have regimes with negative magnetic shear and internal transport barrier to ensure stability at high normalized plasma pressure β{sub N} > 3 (the so-called Advanced scenarios) of interest for FNSF and the commercial reactor. Modeling with full wave three-dimensional codes PSTELION and STELEC2 showed flexible control of the current profile in the reactor plasmas of ITER, FNSF-AT and DEMO [2,3], using multiple frequencies, the positions of the antennae and toroidal waves slow down. Also presented are the results of simulations of current generation by helicons in tokamaks DIII-D, T-15MD and JT-60SA [3]. In DEMO and Power Plant antenna is strongly simplified, being some analoge of mirrors based ECRF launcher, as will be shown. For spherical tokamaks the Helicons excitation scheme does not provide efficient Off-axis CD profile flexibility due to strong coupling of helicons with O-mode, also through the boundary conditions in low aspect machines, and intrinsic large amount of trapped electrons, as is shown by STELION modeling for the NSTX tokamak. Brief history of Helicons experimental and modeling exploration in straight plasmas, tokamaks and tokamak based fusion Reactors projects is given, including planned joint DIII-D – Kurchatov Institute experiment on helicons CD [1].

Fusion - 2050 perspective (in Polish)

CERN Document Server

Romaniuk, R S

2013-01-01

The results of strongly exothermic reaction of thermonuclear fusion between nuclei of deuterium and tritium are: helium nuclei and neutrons, plus considerable kinetic energy of neutrons of over 14 MeV. DT nuclides synthesis reaction is probably not the most favorable one for energy production, but is the most advanced technologically. More efficient would be possibly aneutronic fusion. The EU by its EURATOM agenda prepared a Road Map for research and implementation of Fusion as a commercial method of thermonuclear energy generation in the time horizon of 2050.The milestones on this road are tokomak experiments JET, ITER and DEMO, and neutron experiment IFMIF. There is a hope, that by engagement of the national government, and all research and technical fusion communities, part of this Road Map may be realized in Poland. The infrastructure build for fusion experiments may be also used for material engineering research, chemistry, biomedical, associated with environment protection, power engineering, security, ...

Z-inertial fusion energy: power plant final report FY 2006.

Energy Technology Data Exchange (ETDEWEB)

Anderson, Mark (University of Wisconsin, Madison, WI); Kulcinski, Gerald (University of Wisconsin, Madison, WI); Zhao, Haihua (University of California, Berkeley, CA); Cipiti, Benjamin B.; Olson, Craig Lee; Sierra, Dannelle P.; Meier, Wayne (Lawrence Livermore National Laboratories); McConnell, Paul E.; Ghiaasiaan, M. (Georgia Institute of Technology, Atlanta, GA); Kern, Brian (Georgia Institute of Technology, Atlanta, GA); Tajima, Yu (University of California, Los Angeles, CA); Campen, Chistopher (University of California, Berkeley, CA); Sketchley, Tomas (University of California, Los Angeles, CA); Moir, R (Lawrence Livermore National Laboratories); Bardet, Philippe M. (University of California, Berkeley, CA); Durbin, Samuel; Morrow, Charles W.; Vigil, Virginia L (University of Wisconsin, Madison, WI); Modesto-Beato, Marcos A.; Franklin, James Kenneth (University of California, Berkeley, CA); Smith, James Dean; Ying, Alice (University of California, Los Angeles, CA); Cook, Jason T.; Schmitz, Lothar (University of California, Los Angeles, CA); Abdel-Khalik, S. (Georgia Institute of Technology, Atlanta, GA); Farnum, Cathy Ottinger; Abdou, Mohamed A. (University of California, Los Angeles, CA); Bonazza, Riccardo (University of Wisconsin, Madison, WI); Rodriguez, Salvador B.; Sridharan, Kumar (University of Wisconsin, Madison, WI); Rochau, Gary Eugene; Gudmundson, Jesse (University of Wisconsin, Madison, WI); Peterson, Per F. (University of California, Berkeley, CA); Marriott, Ed (University of Wisconsin, Madison, WI); Oakley, Jason (University of Wisconsin, Madison, WI)

2006-10-01

This report summarizes the work conducted for the Z-inertial fusion energy (Z-IFE) late start Laboratory Directed Research Project. A major area of focus was on creating a roadmap to a z-pinch driven fusion power plant. The roadmap ties ZIFE into the Global Nuclear Energy Partnership (GNEP) initiative through the use of high energy fusion neutrons to burn the actinides of spent fuel waste. Transmutation presents a near term use for Z-IFE technology and will aid in paving the path to fusion energy. The work this year continued to develop the science and engineering needed to support the Z-IFE roadmap. This included plant system and driver cost estimates, recyclable transmission line studies, flibe characterization, reaction chamber design, and shock mitigation techniques.

Z-inertial fusion energy: power plant final report FY 2006

International Nuclear Information System (INIS)

Anderson, Mark; Kulcinski, Gerald; Zhao, Haihua; Cipiti, Benjamin B.; Olson, Craig Lee; Sierra, Dannelle P.; Meier, Wayne; McConnell, Paul E.; Ghiaasiaan, M.; Kern, Brian; Tajima, Yu; Campen, Chistopher; Sketchley, Tomas; Moir, R; Bardet, Philippe M.; Durbin, Samuel; Morrow, Charles W.; Vigil, Virginia L.; Modesto-Beato, Marcos A.; Franklin, James Kenneth; Smith, James Dean; Ying, Alice; Cook, Jason T.; Schmitz, Lothar; Abdel-Khalik, S.; Farnum, Cathy Ottinger; Abdou, Mohamed A.; Bonazza, Riccardo; Rodriguez, Salvador B.; Sridharan, Kumar; Rochau, Gary Eugene; Gudmundson, Jesse; Peterson, Per F.; Marriott, Ed; Oakley, Jason

2006-01-01

This report summarizes the work conducted for the Z-inertial fusion energy (Z-IFE) late start Laboratory Directed Research Project. A major area of focus was on creating a roadmap to a z-pinch driven fusion power plant. The roadmap ties ZIFE into the Global Nuclear Energy Partnership (GNEP) initiative through the use of high energy fusion neutrons to burn the actinides of spent fuel waste. Transmutation presents a near term use for Z-IFE technology and will aid in paving the path to fusion energy. The work this year continued to develop the science and engineering needed to support the Z-IFE roadmap. This included plant system and driver cost estimates, recyclable transmission line studies, flibe characterization, reaction chamber design, and shock mitigation techniques

Numerical analysis of tungsten erosion and deposition processes under a DEMO divertor plasma

Directory of Open Access Journals (Sweden)

Yuki Homma

2017-08-01

Full Text Available Erosion reduction of tungsten (W divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density ni and higher temperature case and the other is higher ni and lower temperature case. Net erosion has been seen in the lower ni case. But in the higher ni case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors are important to understand the net erosion formation: (i ratio of the 1st ionization length of sputtered W atom to the Larmor gyro radius of W+ ion, (ii balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.

Conceptual design of inertial confinement fusion power plant

International Nuclear Information System (INIS)

Mima, Kunioki; Yamanaka, Tatsuhiko; Nakai, Sadao

1994-01-01

Presented is the status of the conceptual design studies of inertial confinement fusion reactors. The recent achievements of the laser fusion research enable us to refine the conceptual design of the power plant. In the paper, main features of several new conceptual designs of ICF reactor; KOYO, SIRIUS-P, HYLIFE-II and so on are summarized. In particular, the target design and the reactor chamber design are described. Finally, the overview of the laser fusion reactor and the irradiation system is also described. (author)

Fusion power core engineering for the ARIES-ST power plant

International Nuclear Information System (INIS)

Tillack, M.S.; Wang, X.R.; Pulsifer, J.; Malang, S.; Sze, D.K.; Billone, M.; Sviatoslavsky, I.

2003-01-01

ARIES-ST is a 1000 MWe fusion power plant based on a low aspect ratio 'spherical torus' (ST) plasma. The ARIES-ST power core was designed to accommodate the unique features of an ST power plant, to meet the top-level requirements of an attractive fusion energy source, and to minimize extrapolation from the fusion technology database under development throughout the world. The result is an advanced helium-cooled ferritic steel blanket with flowing PbLi breeder and tungsten plasma-interactive components. Design improvements, such as the use of SiC inserts in the blanket to extend the outlet coolant temperature range were explored and the results are reported here. In the final design point, the power and particle loads found in ARIES-ST are relatively similar to other advanced tokamak power plants (e.g. ARIES-RS [Fusion Eng. Des. 38 (1997) 3; Fusion Eng. Des. 38 (1997) 87]) such that exotic technologies were not required in order to satisfy all of the design criteria. Najmabadi and the ARIES Team [Fusion Eng. Des. (this issue)] provide an overview of ARIES-ST design. In this article, the details of the power core design are presented together with analysis of the thermal-hydraulic, thermomechanical and materials behavior of in-vessel components. Detailed engineering analysis of ARIES-ST TF and PF systems, nuclear analysis, and safety are given in the companion papers

Design and development of ceramic breeder demo blanket

International Nuclear Information System (INIS)

Enoeda, M.; Sato, S.; Hatano, T.

2001-01-01

Ceramic breeder blanket development has been widely conducted in Japan from fundamental researches to project-oriented engineering scaled development. A long term R and D program has been launched in JAERI since 1996 as a course of DEMO blanket development. The objectives of this program are to provide engineering data base and fabrication technologies of the DEMO blanket, aiming at module testing in ITER currently scheduled to start from the beginning of the ITER operation as a near-term target. Two types of DEMO blanket systems, water cooled blanket and helium cooled blanket, have been designed to be consistent with the SSTR (Steady State Tokamak Reactor) which is the reference DEMO reactor design in JAERI. Both of them utilize packed small pebbles of breeder Li 2 O or Li 2 TiO 3 as a candidate) and neutron multiplier (Be) and rely on the development of advanced structural materials (a reduced activation ferritic steel F82H) compatible with high temperature operation. (author)

Efficiency of water coolant for DEMO divertor

International Nuclear Information System (INIS)

Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

2015-01-01

Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

Efficiency of water coolant for DEMO divertor

Energy Technology Data Exchange (ETDEWEB)

Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

2015-10-15

Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

Consequences of the technology survey and gap analysis on the EU DEMO R&D programme in tritium, matter injection and vacuum

Energy Technology Data Exchange (ETDEWEB)

Day, Chr., E-mail: Christian.Day@kit.edu [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Butler, B. [Culham Science Centre (CCFE), Abingdon (United Kingdom); Giegerich, T. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Lang, P.T. [Max-Planck-Institute of Plasma Physics (IPP), Garching (Germany); Lawless, R. [Culham Science Centre (CCFE), Abingdon (United Kingdom); Meszaros, B. [EUROfusion Consortium, Programme Management Unit, Garching (Germany)

2016-11-01

Highlights: • The inner fuel cycle architecture of DEMO is developed in a systems engineering approach as a functional break-down diagram, driven by the need for inventory minimisation. • Technologies to fulfil the required functions are discussed and ranked. • Prime technologies are identified and an associated R&D programme is developed. • The core challenges of a DEMO fuel cycle beyond those already addressed in ITER are discussed. - Abstract: In the framework of the EUROfusion Programme, EU is preparing the conceptual design of the inner fuel cycle of a pulsed tokamak DEMO. This paper illustrates a quantified process to shape a R&D programme that exploits as much as possible previous R&D. In an initial step, the high-level requirements are collected and a novel DEMO inner fuel cycle architecture with its three sub-systems vacuum pumping, matter injection (fuelling and injection of plasma enhancement gases) and tritium systems (tritium plant and breeder coolant purification) is delineated, driven by the DEMO key challenge to reduce tritium inventory. Then, a technology survey is carried out to review potential existing solutions for the required process functions and to assess their maturity and risks. Finally, a decision-making scheme is applied to select the most promising candidates. ITER technology is exploited where possible. As a primary result, a fuel cycle architecture is suggested with an advanced tritium plant that avoids full isotope separation in the main loop and with a Direct Internal Recycling path in the vacuum systems to shorten cycle times. For core fuelling, classical inboard pellet injection technology is selected, in principle similar to that proposed for ITER but aiming for higher launch speeds to achieve deep fuelling of the DEMO plasma. Based on these findings, a tailored R&D programme is shaped that tackles the key questions until 2020.

Environmental cost/benefit analysis for fusion power plants

International Nuclear Information System (INIS)

Young, J.R.

1976-11-01

This document presents a cost/benefit analysis of use of fusion power plants early in the 21st century. The first section describes the general formulation of the analysis. Included are the selection of the alternatives to the fusion reactor, selection of the power system cases to be compared, and a general comparison of the environmental effects of the selected alternatives. The second section compares the cumulative environmental effects from 2010 to 2040 for the primary cases of the power system with and without fusion reactors. The third section briefly illustrates the potential economic benefits if fusion reactors produce electricity at a lower unit cost than LMFBRs can. The fourth section summarizes the cost/benefit analysis

European structural materials development for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Schaaf, B. van der E-mail: vanderschaaf@nrg-nl.com; Ehrlich, K.; Fenici, P.; Tavassoli, A.A.; Victoria, M

2000-09-01

Leading long term considerations for choices in the European Long Term Technology programme are the high temperature mechanical- and compatibility properties of structural materials under neutron irradiation. The degrees of fabrication process freedom are closely investigated to allow the construction of complex shapes. Another important consideration is the activation behaviour of the structural material. The ideal solution is the recycling of the structural materials after a relatively short 'cooling' period. The structural materials development in Europe has three streams. The first serves the design and construction of ITER and is closely connected to the choice made: water cooled austenitic stainless steel. The second development stream is to support the design and construction of DEMO relevant blanket modules to be tested in ITER. The helium cooled pebble bed and the water cooled liquid lithium concept rely both on RAFM steel. The goal of the third stream is to investigate the potential of advanced materials for fusion power reactors beyond DEMO. The major contending materials: SiCSiC composites, vanadium, titanium and chromium alloys hold the promise of high operating temperatures, but RAFM has also a high temperature potential applying oxide dispersion strengthening. The development of materials for fusion power application requires a high flux 14 MeV neutron source to simulate the fusion power environment.

Demo of Gaze Controlled Flying

DEFF Research Database (Denmark)

Alapetite, Alexandre; Hansen, John Paulin; Scott MacKenzie, I.

2012-01-01

Development of a control paradigm for unmanned aerial vehicles (UAV) is a new challenge to HCI. The demo explores how to use gaze as input for locomotion in 3D. A low-cost drone will be controlled by tracking user’s point of regard (gaze) on a live video stream from the UAV.......Development of a control paradigm for unmanned aerial vehicles (UAV) is a new challenge to HCI. The demo explores how to use gaze as input for locomotion in 3D. A low-cost drone will be controlled by tracking user’s point of regard (gaze) on a live video stream from the UAV....

Analysis of the WCLL European demo blanket concept in terms of activation and decay heat after exposure to neutron irradiation

OpenAIRE

Stankunas Gediminas; Tidikas Andrius

2017-01-01

This comparative paper describes the activation and decay heat calculations for water-cooled lithium-lead performed part of the EURO fusion WPSAE programme and specifications in comparison to other European DEMO blanket concepts on the basis of using a three-dimensional neutronics calculation model. Results are provided for a range of decay times of interest for maintenance activities, safety and waste management assessments. The study revealed that water-c...

Development of a master model concept for DEMO vacuum vessel

International Nuclear Information System (INIS)

Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea; Bachmann, Christian; Di Gironimo, Giuseppe

2016-01-01

Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

Development of a master model concept for DEMO vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Mozzillo, Rocco; Marzullo, Domenico; Tarallo, Andrea [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy)

2016-11-15

Highlights: • The present work concerns the development of a first master concept model for DEMO vacuum vessel. • A parametric-associative CAD master model concept of a DEMO VV sector has been developed in accordance with DEMO design guidelines. • A proper CAD design methodology has been implemented in view of the later FEM analyses based on “shell elements”. - Abstract: This paper describes the development of a master model concept of the DEMO vacuum vessel (VV) conducted within the framework of the EUROfusion Consortium. Starting from the VV space envelope defined in the DEMO baseline design 2014, the layout of the VV structure was preliminarily defined according to the design criteria provided in RCC-MRx. A surface modelling technique was adopted and efficiently linked to the finite element (FE) code to simplify future FE analyses. In view of possible changes to shape and structure during the conceptual design activities, a parametric design approach allows incorporating modifications to the model efficiently.

Multi-unit Inertial Fusion Energy (IFE) plants producing hydrogen fuel

International Nuclear Information System (INIS)

Logan, B.G.

1993-12-01

A quantitative energy pathway comparison is made between a modern oil refinery and genetic fusion hydrogen plant supporting hybrid-electric cars powered by gasoline and hydrogen-optimized internal combustion engines, respectively, both meeting President Clinton's goal for advanced car goal of 80 mpg gasoline equivalent. The comparison shows that a fusion electric plant producing hydrogen by water electrolysis at 80% efficiency must have an electric capacity of 10 GWe to support as many hydrogen-powered hybrid cars as one modern 200,000 bbl/day-capacity oil refinery could support in gasoline-powered hybrid cars. A 10 GWe fusion electric plant capital cost is limited to 12.5 B$ to produce electricity at 2.3 cents/kWehr, and hydrogen production by electrolysis at 8 $/GJ, for equal consumer fuel cost per passenger mile as in the oil-gasoline-hybrid pathway

Breeding blanket for Demo

International Nuclear Information System (INIS)

Proust, E.; Giancarli, L.

1992-01-01

This paper presents the main design features, their rationale, and the main critical issues for the development, of the four DEMO-relevant blanket concepts presently investigated within the framework of the European Test-Blanket Development Programme

Progress of R&D on water cooled ceramic breeder for ITER test blanket system and DEMO

Energy Technology Data Exchange (ETDEWEB)

Kawamura, Yoshinori, E-mail: kawamura.yoshinori@jaea.go.jp [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Tanigawa, Hisashi; Hirose, Takanori; Enoeda, Mikio [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Sato, Satoshi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Konno, Chikara; Edao, Yuki; Hayashi, Takumi [Japan Atomic Energy Agency, 2-4 Shirane Shirakata, Tokai, Ibaraki 319-1195 (Japan); Hoshino, Tsuyoshi; Nakamichi, Masaru; Tanigawa, Hiroyasu [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan); Nishi, Hiroshi; Suzuki, Satoshi; Ezato, Koichiro; Seki, Yohji [Japan Atomic Energy Agency, 801-1 Mukoyama, Naka, Ibaraki 311-0193 (Japan); Yamanishi, Toshihiko [Japan Atomic Energy Agency, 2-166 Omotedate Obuchi, Rokkasho, Aomori 039-3212 (Japan)

2016-11-01

Highlights: • Thermo-hydraulic calculation in the TBM at the water ingress event has been done. • Shielding calculations for the ITER equatorial port #18 were conducted by using C-lite model. • Prototypic pebbles of Be{sub 17}Ti{sub 2} and Be{sub 12}V had a good oxidation property similar to Be{sub 12}Ti pebble. • Li rich Li{sub 2}TiO{sub 3} pebbles were successfully fabricated using the emulsion method by controlling sintering atmosphere. • New tritium production/recovery experiments at FNS have been started by using ionization chamber as on-line gas monitor. - Abstract: The development of a water cooled ceramic breeder (WCCB) test blanket module (TBM) is being performed as one of the most important steps toward DEMO blanket in Japan. For the TBM testing and development of DEMO blanket, R&D has been performed on the module fabrication technology, breeder and multiplier pebble fabrication technology, tritium production rate evaluation, as well as structural and safety design activities. The fabrication of full-scale first wall, side walls, breeder pebble bed box and back wall was completed, and assembly of TBM with box structure was successfully achieved. Development of advanced breeder and multiplier pebbles for higher chemical stability was continued for future DEMO blanket application. From the view point of TBM test result evaluation and DEMO blanket performance design, the development of the blanket tritium transport simulation technology, investigation of the TBM neutron measurement technology and the evaluation of the tritium production and recovery test using D-T neutron in the fusion neutron source (FNS) facility has been performed. This paper provides an overview of the recent achievements of the development of the WCCB Blanket in Japan.

Neutronic analysis of a dual He/LiPb coolant breeding blanket for DEMO

International Nuclear Information System (INIS)

Catalan, J.P.; Ogando, F.; Sanz, J.; Palermo, I.; Veredas, G.; Gomez-Ros, J.M.; Sedano, L.

2011-01-01

A conceptual design of a DEMO fusion reactor is being developed under the Spanish Breeding Blanket Technology Programme: TECNO F US based on a He/LiPb dual coolant blanket as reference design option. The following issues have been analyzed to address the demonstration of the neutronic reliability of this conceptual blanket design: power amplification capacity of the blanket, tritium breeding capability for fuel self-sufficiency, power deposition due to nuclear heating in superconducting coils and material damage (dpa, gas production) to estimate the operational life of the steel-made structural components in the blanket and vacuum vessel (VV). In order to optimize the shielding of the coils different combinations of water and steel have been considered for the gap of the VV. The used neutron source is based on an axi-symmetric 2D fusion reaction profile for the given plasma equilibrium configuration. MCNPX has been used for transport calculations and ACAB has been used to handle gas production and damage energy cross sections.

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 1

International Nuclear Information System (INIS)

Malang, S.; Reimann, J.; Sebening, H.; Barleon, L.; Bogusch, E.; Bojarsky, E.; Borgstedt, H.U.; Buehler, L.; Casal, V.; Deckers, H.; Feuerstein, H.; Fischer, U.; Frees, G.; Graebner, H.; John, H.; Jordan, T.; Kramer, W.; Krieg, R.; Lenhart, L.; Malang, S.; Meyder, R.; Norajitra, P.; Reimann, J.; Schwenk-Ferrero, A.; Schnauder, H.; Stieglitz, R.; Oschinski, J.; Wiegner, E.

1991-12-01

A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary, Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated R and D-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required R and D-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

An assessment for the erosion rate of DEMO first wall

Science.gov (United States)

Tokar, M. Z.

2018-01-01

In a fusion reactor a significant fraction of plasma particles lost from the confined volume will reach the vessel wall. The recombination of these charged species, electrons and ions of hydrogen isotopes, is a source of neutral molecules and atoms, recycling back into the plasma. Here they participate, in particular, in charge-exchange (c-x) collisions with the plasma ions and, as a result, atoms of high energies with chaotically oriented velocities are generated. A significant fraction of these hot neutrals will hit the wall, leading, as well as the outflowing fuel and impurity ions, to its erosion, limiting the reactor operation time. The rate of the wall erosion in DEMO is assessed by applying a one-dimensional model which takes into account the transport of charged and neutral species across the flux surfaces in the main part of the scrape-off layer, beyond the X-point vicinity and divertor, and by considering the shift of the centers of flux surfaces, their elongation and triangularity. Atoms generated by c-x of recycling neutrals are modeled kinetically to define firmly their energy spectrum, being of particular importance for the erosion assessment. It is demonstrated the erosion rate of the DEMO wall armor of tungsten will have a pronounced ballooning character with a significant maximum of 0.3 mm per full power year at the low field side, decreasing with an increase in the anomalous perpendicular transport in the ‘far’ SOL or the plasma density at the separatrix.

Potential environmental effects of fusion reactor power plants

International Nuclear Information System (INIS)

Young, J.R.; Gore, B.F.; Coffman, F.E.

1976-01-01

Construction and operation of fusion power plants is expected to reduce the total environmental effects of 21st century power generation. Fusion power plant impacts due to noise, odors, vibrations, and sanitary wastes are expected to be insignificant. impacts due to land use, chemical releases, and aesthetics are expected to be reduced. Impacts due to heat releases, local socio-economic changes, and non-radioactive liquid and solid disposal are expected to be comparable to those for the alternative fission or coal-fired power systems. Radiation doses to the public due to radioactive wastes are expected to be comparable to, or less than, the trivial low doses due to fission power systems. Research and development will be required, however, to assure adequate containment of tritium, the primary radioisotope of concern. Prevention of accidental tritium releases is within the capability of current engineering practice. Current technology is capable of handling the solid radioactive waste which may be produced, with insignificant environmental impact. Major research efforts are necessary to determine if subtle long-term effects of magnetic fields exist and should be of concern. In view of the large quantities of construction materials required for fusion. Material availability may dictate 21st century power plant design and construction. The accident potential of fusion power plants should be lower than for fission systems. Accidental criticalities and plasma runaways are not considered to be possible. Loss of coolant accidents are not expected to result in damage to the containment. No fission products or actinides are present to be released in an accident, and most activation products are immobilized in structures. The biological hazard of tritium is orders of magnitude smaller than for fission products and actinides. Safeguards against diversion of fissile materials are not expected to be necessary

Structural materials requirements for in-vessel components of fusion power plants

International Nuclear Information System (INIS)

Schaaf, B. van der

2000-01-01

The economic production of fusion energy is determined by principal choices such as using magnetic plasma confinement or generating inertial fusion energy. The first generation power plants will use deuterium and tritium mixtures as fuel, producing large amounts of highly energetic neutrons resulting in radiation damage in materials. In the far future the advanced fuels, 3 He or 11 B, determine power plant designs with less radiation damage than in the first generation. The first generation power plants design must anticipate radiation damage. Solid sacrificing armour or liquid layers could limit component replacements costs to economic levels. There is more than radiation damage resistance to determine the successful application of structural materials. High endurance against cyclic loading is a prominent requirement, both for magnetic and inertial fusion energy power plants. For high efficiency and compactness of the plant, elevated temperature behaviour should be attractive. Safety and environmental requirements demand that materials have low activation potential and little toxic effects under both normal and accident conditions. The long-term contenders for fusion power plant components near the plasma are materials in the range from innovative steels, such as reduced activation ferritic martensitic steels, to highly advanced ceramic composites based on silicon carbide, and chromium alloys. The steels follow an evolutionary path to basic plant efficiencies. The competition on the energy market in the middle of the next century might necessitate the riskier but more rewarding development of SiCSiC composites or chromium alloys

Data fusion and sensor management for nuclear power plant safety

International Nuclear Information System (INIS)

Ciftcioglu, Oe.

1996-05-01

The paper describes the implementation of the data-sensor fusion and sensor management technology for accident management through simulated severe accident (SA) scenarios subjected to study. By means of accident management the appropriate prompt actions to be taken to avoid nuclear accident (SA) scenarios subjected to study. By means of accident management the appropriate prompt actions to be taken to avoid nuclear accidents are meant, while such accidents are deemed to somehow be imminent during plant operation. The organisation of the present paper is as follows. As the data-sensor fusion and sensor management is an emerging technology which is not widely known, in Sec. 2, the definition and goals of data-sensor fusion and sensor management technology is described. In Sec. 3 first, with reference to Kalman filtering as an information filter, statistical data-sensor fusion technology is described. This is followed by the examples of deterministic data-sensor fusion technology using gross plant state variables and neural networks (NN) and the implementation for severe accident management in NPPs. In Sec. 4, the sensor management technology is described. Finally, the performance of the data-sensor fusion technology for NPP safety is discussed. (orig./WL)

The DEMO Quasisymmetric Stellarator

Directory of Open Access Journals (Sweden)

Geoffrey B. McFadden

2010-02-01

Full Text Available The NSTAB nonlinear stability code solves differential equations in conservation form, and the TRAN Monte Carlo test particle code tracks guiding center orbits in a fixed background, to provide simulations of equilibrium, stability, and transport in tokamaks and stellarators. These codes are well correlated with experimental observations and have been validated by convergence studies. Bifurcated 3D solutions of the 2D tokamak problem have been calculated that model persistent disruptions, neoclassical tearing modes (NTMs and edge localized modes (ELMs occurring in the International Thermonuclear Experimental Reactor (ITER, which does not pass the NSTAB simulation test for nonlinear stability. So we have designed a quasiaxially symmetric (QAS stellarator with similar proportions as a candidate for the demonstration (DEMO fusion reactor that does pass the test [1]. The configuration has two field periods and an exceptionally accurate 2D symmetry that furnishes excellent thermal confinement and good control of the prompt loss of alpha particles. Robust coils are found from a filtered form of the Biot-Savart law based on a distribution of current over a control surface for the coils and the current in the plasma defined by the equilibrium calculation. Computational science has addressed the issues of equilibrium, stability, and transport, so it remains to develop an effective plan to construct the coils and build a diverter.

Parameters of DEMO DN and JET DN

International Nuclear Information System (INIS)

Anon.

1987-01-01

The paper is an appendix to a study of the reactor relevance of the NET design concept. The latter study examines whether the technologies and design principles proposed for NET can be directly extrapolated to a demonstration (DEMO) reactor. The appendix presents the parameters of the DEMO and NET under the topic headings: power, geometry, plasma, toroidal and poloidal magnetic field coils, first wall engineering, divertor physics, divertor engineering, and blanket. (U.K.)

Nuclear Power Plants Fault Diagnosis Method Based on Data Fusion

International Nuclear Information System (INIS)

Xie Chunli; Liu Yongkuo; Xia Hong

2009-01-01

The data fusion is a method suit for complex system fault diagnosis such as nuclear power plants, which is multisource information processing technology. This paper uses data fusion information hierarchical thinking and divides nuclear power plants fault diagnosis into three levels. Data level adopts data mining method to handle data and reduction attributes. Feature level uses three parallel neural networks to deal with attributes of data level reduction and the outputs of three networks are as the basic probability assignment of Dempster-Shafer (D-S) evidence theory. The improved D-S evidence theory synthesizes the outputs of neural networks in decision level, which conquer the traditional D-S evidence theory limitation which can't dispose conflict information. The diagnosis method was tested using correlation data of literature. The test results indicate that the data fusion diagnosis system can diagnose nuclear power plants faults accurately and the method has application value. (authors)

Management of nontritium radioactive wastes from fusion power plants

International Nuclear Information System (INIS)

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

1976-09-01

This report identifies nontritium radioactive waste sources for current conceptual fusion reactor designs. Quantities and compositions of the radwaste are estimated for the tokamaks of the University of Wisconsin (UWMAK-I), the Princeton Plasma Physics Laboratory (PPPL), and the Oak Ridge National Laboratory (ORNL); the Reference Theta Pinch Reactor of the Los Alamos Scientific Laboratory (LASL); and the Minimum Activation Blanket of the Brookhaven National Laboratory (BNL). Disposal of large amounts of radioactive waste appears necessary for fusion reactors. Although the curie (Ci) level of the wastes is comparable to that of fission products in fission reactors, the isotopes are less hazardous, and have shorter half-lives. Therefore radioactivity from fusion power production should pose a smaller risk than radioactivity from fission reactors. Radioactive waste sources identified for the five reference plants are summarized. Specific radwaste treatments or systems had to be assumed to estimate these waste quantities. Future fusion power plant conceptual designs should include radwaste treatment system designs so that assumed designs do not have to be used to assess the environmental effects of the radioactive waste

Recycling of copper used in fusion power plants

International Nuclear Information System (INIS)

Forty, C.B.A.; Butterworth, G.J.; Turner, A.D.; Junkison, A.J.

1997-04-01

One of the major safety and environmental advantages of fusion power is a limited waste management burden on future generations. In this connection, the ability to recycle end-of-service materials from fusion power plant is beneficial both in terms of the conservation of natural resources and the minimisation of the volumes of activated wastes. After 100 years, the residual activity of near-plasma copper components exceeds that permitted for free release or contact handling. The presence of silver as a common impurity in copper may exacerbate this problem, through generation of 108m Ag. Removal of the silver impurity in a separate refining step prior to use of the copper in a fusion plant obviates the problems associated with formation of 108m Ag. Two alternative desilveration processes have been demonstrated; one involving the segregation of silver as AgBr and the other the absorption of Ag + by ion exchange. The present study demonstrates that conventional electrorefining techniques can be adapted to recover used copper in a single refining stage, with sufficient decontamination to permit its reuse in fusion power plants or, with a second stage, unrestricted release. Shielding requirements for the processing of scrap copper in conventional hot cells indicate a decay storage period of 50-100 years. To maximise the cost of savings of reclamation over direct geological disposal, the activation products may be separated out and disposed of in a metallic form. A substantial reduction in the overall volume of active waste should thus be achievable, especially if supercompaction can be applied to the product. (Author)

Recycling of copper used in fusion power plants

International Nuclear Information System (INIS)

Butterworth, G.J.; Forty, C.B.A.

1998-01-01

One of the major safety and environmental advantages of fusion power is a limited waste management burden on future generations. In this connection, the ability to recycle end-of-service materials from fusion power plants is beneficial both in terms of the conservation of natural resources and the minimisation of the volume of activated wastes. After 100 years, the residual activity of near-plasma copper components exceeds that permitted for free release or contact handling. The presence of silver as a common impurity in copper may exacerbate this problem, through generation of 108m Ag. Removal of the silver impurity in a separate refining step prior to use of the copper in a fusion plant obviates the problems associated with formation of 108m Ag. Two alternative desilverisation processes have been demonstrated; one involving the segregation of silver as AgBr and the other the absorption of Ag + by ion exchange. The present study demonstrates that conventional electrorefining techniques can be adapted to recover used copper in a single refining stage, with sufficient decontamination to permit its reuse in fusion power plants or, with a second stage, unrestricted release. Shielding requirements for the processing of scrap copper in conventional hot cells indicate a decay storage period of 50-100 years. To maximise the cost savings of reclamation over direct geological disposal, the activation products may be separated out and disposed of in a metallic form. A substantial reduction in the overall volume of active waste should thus be achievable, especially if supercompaction can be applied to the product. (orig.)

CORIANDER: Comparison of relevant issues and nuclear development for fusion energy research

International Nuclear Information System (INIS)

Arendt, F.; Boehme, G.; Haferkamp, B.; Heinz, W.; Hutter, E.; Klingelhoefer, R.; Komarek, P.; Kramer, W.; Kuechle, M.; Kuntze, M.; Malang, S.; Maurer, W.; Mueller, R.A.; Perinic, D.; Leppelmeier, G.; El-Guebaly, L.A.; Emmert, G.A.; Kulcinski, G.L.; Larsen, E.M.; Maynard, C.W.; Santarius, J.F.; Sawan, M.E.; Scharer, J.E.; Sviatoslavsky, I.N.; Vogelsang, W.F.; Wittenberg, L.J.

1985-05-01

This study compares two strategies to be followed on the way to a tokamak demonstration reactor (DEMO). The first is the present European conception of building an integrated physics and technology machine (NET) between the present large physics experiments (such as JET) and DEMO. In the Alternate Plan, NET woild be replaced by a combination of an advanced physics tokamak and a mirror based dedicated fusion technology device ('TASKA-class'). It appears that the Alternate Plan could provide the required physics and most of the engineering data for building DEMO with less risk, in a shorter time, and perhaps less cost than the present route. While it is highly desirable to increase the neutron fluence for material studies in either strategy, the Alternat Plan may require extrapolations in the combination of blanket geometry effects and high neutron fluence. (orig.) [de

Neutrons and fusion nuclear technology

International Nuclear Information System (INIS)

Hirayama, Shoichi

1991-01-01

The strategy of the devolopment of the fusion reactor has been compared with the history of the development of the fission reactor. More than 50 neutron reactors (neutron sources for research and development of reactor components and materials, and for Pu production) have been constructed and operated before the introduction of demonstration power reactors. This fact suggests us to introduce a new path of neutron reactor in the strategy of the development of fusion power reactor in addition to the orthodox approach which goes through the break-even, self-ignition, ETR, and DEMO. One of the benefits of the introduction of such neutron reactor or into the strategy of the fusion reactor development has been studied numerically. The results demonstrate that the introduction of fission-fusion hybrid reactor in 2030, can save ∝20% of natural uranium by 2100 in Japan, in comparison with the case when the fast breeder reactor is introduced in 2030. This saving is recognized large enough to justify earlier construction of the fusion neutron reactor. (orig.)

FRESCO, a simplified code for cost analysis of fusion power plants

International Nuclear Information System (INIS)

Bustreo, C.; Casini, G.; Zollino, G.; Bolzonella, T.; Piovan, R.

2013-01-01

Highlights: • FRESCO is a code for rapid evaluation of the cost of electricity of a fusion power plant. • Parameters of the basic machine and unitary costs of components derived from ITER. • Power production components and plant power balance are extrapolated from PPCS. • A special effort is made in the investigation of the pulsed operation scenarios. • Technical and economical FRESCO results are compared with those of two PPCS models. -- Abstract: FRESCO (Fusion REactor Simplified COsts) is a code based on simplified models of physics, engineering and economical aspects of a TOKAMAK-like pulsed or steady-state fusion power plant. The experience coming from various aspects of ITER design, including selection of materials and operating scenarios, is exploited as much as possible. Energy production and plant power balance, including the recirculation requirements, are derived from two models of the PPCS European study, the helium cooled lithium/lead blanket model reactor (model AB) and the helium cooled ceramic one (model B). A detailed study of the availability of the power plant due, among others, to the replacement of plasma facing components, is also included in the code. The economics of the fusion power plant is evaluated through the levelized cost approach. Costs of the basic components are scaled from the corresponding values of the ITER project, the ARIES studies and SCAN model. The costs of plant auxiliaries, including those of the magnetic and electric systems, tritium plants, instrumentation, buildings and thermal energy storage if any, are recovered from ITER values and from those of other power plants. Finally, the PPCS models AB and B are simulated and the main results are reported in this paper

Tritium extraction technologies and DEMO requirements

Energy Technology Data Exchange (ETDEWEB)

Demange, D., E-mail: david.demange@kit.edu [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Antunes, R.; Borisevich, O.; Frances, L. [Karlsruhe Institute of Technology, Institute for Technical Physics, Tritium Laboratory Karlsruhe, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rapisarda, D. [Laboratorio Nacional de Fusión, EURATOM-CIEMAT, 28040 Madrid (Spain); Santucci, A. [ENEA for EUROfusion, Via E. Fermi 45, 00044 Frascati, Roma (Italy); Utili, M. [ENEA CR Brasimone, 40032 Camugnano, BO (Italy)

2016-11-01

Highlights: • We detail the R&D plan for tritium technology of the European DEMO breeding blanket. • We study advanced and efficient extraction techniques to improve tritium management. • We consider inorganic membranes and catalytic membrane reactor for solid blankets. • We consider permeator against vacuum and vacuum sieve tray for liquid blankets. - Abstract: The conceptual design of the tritium extraction system (TES) for the European DEMO reactor is worked out in parallel for four different breeding blankets (BB) retained by EUROfusion. The TES design has to be tackled in an integrated manner optimizing the synergy with the directly interfacing inner fuel cycle, while minimizing the tritium permeation into the coolant. Considering DEMO requirements, it is most likely that only advanced technologies will be suitable for the tritium extraction systems of the BB. This paper overviews the European work programme for R&D on tritium technology for the DEMO BB, summaries the general first outcomes, and details the specific and comprehensive R&D program to study experimentally immature but promising technologies such as vacuum sieve tray or permeator against vacuum for tritium extraction from PbLi, and advanced inorganic membranes and catalytic membrane reactor for tritium extraction from He. These techniques are simple, fully continuous, likely compact with contained energy consumption. Several European Laboratories are joining their efforts to deploy several new experimental setups to accommodate the tests campaigns that will cover small scale experiments with tritium and inactive medium scale tests so as to improve the technology readiness level of these advanced processes.

Massively Clustered CubeSats NCPS Demo Mission

Science.gov (United States)

Robertson, Glen A.; Young, David; Kim, Tony; Houts, Mike

2013-01-01

Technologies under development for the proposed Nuclear Cryogenic Propulsion Stage (NCPS) will require an un-crewed demonstration mission before they can be flight qualified over distances and time frames representative of a crewed Mars mission. In this paper, we describe a Massively Clustered CubeSats platform, possibly comprising hundreds of CubeSats, as the main payload of the NCPS demo mission. This platform would enable a mechanism for cost savings for the demo mission through shared support between NASA and other government agencies as well as leveraged commercial aerospace and academic community involvement. We believe a Massively Clustered CubeSats platform should be an obvious first choice for the NCPS demo mission when one considers that cost and risk of the payload can be spread across many CubeSat customers and that the NCPS demo mission can capitalize on using CubeSats developed by others for its own instrumentation needs. Moreover, a demo mission of the NCPS offers an unprecedented opportunity to invigorate the public on a global scale through direct individual participation coordinated through a web-based collaboration engine. The platform we describe would be capable of delivering CubeSats at various locations along a trajectory toward the primary mission destination, in this case Mars, permitting a variety of potential CubeSat-specific missions. Cameras on various CubeSats can also be used to provide multiple views of the space environment and the NCPS vehicle for video monitoring as well as allow the public to "ride along" as virtual passengers on the mission. This collaborative approach could even initiate a brand new Science, Technology, Engineering and Math (STEM) program for launching student developed CubeSat payloads beyond Low Earth Orbit (LEO) on future deep space technology qualification missions. Keywords: Nuclear Propulsion, NCPS, SLS, Mars, CubeSat.

Environmental life cycle assessment of high temperature nuclear fission and fusion biomass gasification plants

International Nuclear Information System (INIS)

Takeda, Shutaro; Sakurai, Shigeki; Kasada, Ryuta; Konishi, Satoshi

2017-01-01

The authors propose nuclear biomass gasification plant as an advancement of conventional gasification plants. Environmental impacts of both fission and fusion plants were assessed through life cycle assessment. The result suggested the reduction of green-house gas emissions would be as large as 85.9% from conventional plants, showing a potential for the sustainable future for both fission and fusion plants. (author)

Osiris and SOMBRERO inertial confinement fusion power plant designs

International Nuclear Information System (INIS)

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

Conceptual designs and assessments have been completed for two inertial fusion energy (IFE) electric power plants. The detailed designs and results of the assessment studies are presented in this report. Osiris is a heavy-ion-beam (HIB) driven power plant and SOMBRERO is a Krypton-Fluoride (KrF) laser-driven power plant. Both plants are sized for a net electric power of 1000 MWe

Science assessment of fusion power plant

International Nuclear Information System (INIS)

Nagai, Toru; Shimazu, Yasuo

1984-01-01

A concept of SCIENCE ASSESSMENT (SA) is proposed to support a research program of the so-called big science. The SA System should be established before the demonstration reactor is realized, and the system is classified into four categories: (1) Resource Economy Assessment (REA) (cost evaluation and availability of rare resource materials), (2) Risk Assessment (RA) (structural safety during operation and accident), (3) Environmental Assessment (EA) (adaptability to environments), and (4) Socio-Political Assessment (SPA) (from local public acceptance to national policy acceptance). Here, REA to the published conceptual designs of commercial fusion power plants (most of them are TOKAMAK) is carried out as the first step. The energy analysis method is imployed because the final goal of fusion plant is to supply energy. The evaluation index is the energy ratio (= output/input). Computer code for energy analysis was developed, to which the material inventory table from the conceptual design and the database for the energy intensity (= energy required to obtain a unit amount of materials) were prepared. (Nogami, K.)

High temperature superconductors for fusion magnets -influence of neutron irradiation

International Nuclear Information System (INIS)

Chudy, M.; Eisterer, M.; Weber, H. W.

2010-01-01

In this work authors present the results of study of influence of neutron irradiation of high temperature superconductors for fusion magnets. High temperature superconductors (type of YBCO (Yttrium-Barium-Copper-Oxygen)) are strong candidates to be applied in the next step of fusion devices. Defects induced by fast neutrons are effective pinning centres, which can significantly improve critical current densities and reduce J c anisotropy. Due to induced lattice disorder, T c is reduced. Requirements for ITER (DEMO) are partially achieved at 64 K.

Fusion technology programme

International Nuclear Information System (INIS)

Finken, D.

1984-10-01

The KfK-Association has continued work on 17 R and D contracts of the Fusion Technology Programme. An effort of 94 manyears per year is at present contributed by 10 KfK departments, covering all aereas defined in the Fusion Technology Programme. The dominant part of the work is directed towards the need of the NET design or supporting experiments. Some additional effort addresses long term technological issues and system studies relevant to DEMO or confinement schemes alternative to tokamaks. Direct contribution to the NET team has increased by augmentation of NET study contracts and delegation of personnel, three KfK delegates being at present members of the NET team. In reverse, specifications and design guidelines worked out by NET have started to have an impact on the current R and D-work in the laboratory. (orig./GG)

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 1: Self-cooled liquid metal breeder blanket. Vol. 2. Detailed version

International Nuclear Information System (INIS)

John, H.; Malang, S.; Sebening, H.

1991-12-01

A self-cooled liquid metal breeder blanket for a fusion DEMO-reactor and the status of the development programme is described as a part of the European development programme of DEMO relevant test blankets for NET/ITER. Volume 1 (KfK 4907) contains a summary. Volume 2 (KfK 4908) a more detailed version of the report. Both volumes contain sections on previous studies on self-cooled liquid metal breeder blankets, the reference blanket design for a DEMO-reactor, a typical test blanket design including the ancillary loop system and the building requirements for NET/ITER together with the present status of the associated RandD-programme in the fields of neutronics, magnetohydrodynamics, tritium removal and recovery, liquid metal compatibility and purification, ancillary loop system, safety and reliability. An outlook is given regarding the required RandD-programme for the self-cooled liquid metal breeder blanket prior to tests in NET/ITER and the relevant test programme to be performed in NET/ITER. (orig.) [de

Thermal-hydraulic analysis of water cooled breeding blanket of K-DEMO using MARS-KS code

Energy Technology Data Exchange (ETDEWEB)

Lee, Jeong-Hun; Park, Il Woong; Kim, Geon-Woo; Park, Goon-Cherl [Seoul National University, Seoul (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Seoul National University, Seoul (Korea, Republic of); Im, Kihak [National Fusion Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Highlights: • The thermal design of breeding blanket for the K-DEMO is evaluated using MARS-KS. • To confirm the prediction capability of MARS, the results were compared with the CFD. • The results of MARS-KS calculation and CFD prediction are in good agreement. • A transient simulation was carried out so as to show the applicability of MARS-KS. • A methodology to simulate the entire blanket system is proposed. - Abstract: The thermal design of a breeding blanket for the Korean Fusion DEMOnstration reactor (K-DEMO) is evaluated using the Multidimensional Analysis of Reactor Safety (MARS-KS) code in this study. The MARS-KS code has advantages in simulating transient two-phase flow over computational fluid dynamics (CFD) codes. In order to confirm the prediction capability of the code for the present blanket system, the calculation results were compared with the CFD prediction. The results of MARS-KS calculation and CFD prediction are in good agreement. Afterwards, a transient simulation for a conceptual problem was carried out so as to show the applicability of MARS-KS for a transient or accident condition. Finally, a methodology to simulate the multiple blanket modules is proposed.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

The concept of system for chips production need to work demo CHP plant in company 'AGROSAVA' from Šimanovci

Directory of Open Access Journals (Sweden)

Dedić Aleksandar Đ.

2014-01-01

Full Text Available In this paper according to the calculation of chips productivity needs for gasification in the demo CHP plant for co-generation: electricity and heat, chippers were analyzed due to: the type of mobility, running for chipping and the method of delivering chips to temporary yard. The plant was planned to generate electricity power up to 200kWelec. First, in consideration were taken the chippers with medium capacity, which mainly served for chipping brushwood and leaves that remain after harvest plantations on mostly flat terrain and parks. Later, the comparative characteristics of the world's three largest manufacturers of machinery for the production of wood chips significantly larger amounts (up to 30m3/h were given. These chippers were particularly suitable for the higher density of crops and stationed yard, in which brushwood would be brought and chip. At the end, the types of convective dryers were analyzed that could be successfully used for drying wood chips (drum and pneumatic dryer and based on the calculation proposed the types of dryers that were available in the local market.

Control of a laser inertial confinement fusion-fission power plant

Science.gov (United States)

Moses, Edward I.; Latkowski, Jeffery F.; Kramer, Kevin J.

2015-10-27

A laser inertial-confinement fusion-fission energy power plant is described. The fusion-fission hybrid system uses inertial confinement fusion to produce neutrons from a fusion reaction of deuterium and tritium. The fusion neutrons drive a sub-critical blanket of fissile or fertile fuel. A coolant circulated through the fuel extracts heat from the fuel that is used to generate electricity. The inertial confinement fusion reaction can be implemented using central hot spot or fast ignition fusion, and direct or indirect drive. The fusion neutrons result in ultra-deep burn-up of the fuel in the fission blanket, thus enabling the burning of nuclear waste. Fuels include depleted uranium, natural uranium, enriched uranium, spent nuclear fuel, thorium, and weapons grade plutonium. LIFE engines can meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the highly undesirable stockpiles of depleted uranium, spent nuclear fuel and excess weapons materials.

Design study of ITER-like divertor target for DEMO

International Nuclear Information System (INIS)

Crescenzi, Fabio; Bachmann, C.; Richou, M.; Roccella, S.; Visca, E.; You, J.-H.

2015-01-01

Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m"−"2, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

Design study of ITER-like divertor target for DEMO

Energy Technology Data Exchange (ETDEWEB)

Crescenzi, Fabio, E-mail: fabio.crescenzi@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bachmann, C. [EFDA, Power Plant Physics and Technology, Boltzmannstraße 2, 85748 Garching (Germany); Richou, M. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Roccella, S.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); You, J.-H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

2015-10-15

Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m{sup −2}, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

Systems Modeling For The Laser Fusion-Fission Energy (LIFE) Power Plant

International Nuclear Information System (INIS)

Meier, W.R.; Abbott, R.; Beach, R.; Blink, J.; Caird, J.; Erlandson, A.; Farmer, J.; Halsey, W.; Ladran, T.; Latkowski, J.; MacIntyre, A.; Miles, R.; Storm, E.

2008-01-01

A systems model has been developed for the Laser Inertial Fusion-Fission Energy (LIFE) power plant. It combines cost-performance scaling models for the major subsystems of the plant including the laser, inertial fusion target factory, engine (i.e., the chamber including the fission and tritium breeding blankets), energy conversion systems and balance of plant. The LIFE plant model is being used to evaluate design trade-offs and to identify high-leverage R and D. At this point, we are focused more on doing self consistent design trades and optimization as opposed to trying to predict a cost of electricity with a high degree of certainty. Key results show the advantage of large scale (>1000 MWe) plants and the importance of minimizing the cost of diodes and balance of plant cost

STAR Power, an Interactive Educational Fusion CD with a Dynamic, Shaped Tokamak Power Plant Simulator

Science.gov (United States)

Leuer, J. A.; Lee, R. L.; Kellman, A. G.; Chapman Nutt, G. C., Jr.; Holley, G.; Larsen, T. A.

2000-10-01

We describe an interactive, educational fusion adventure game developed within our fusion education program. The theme of the adventure is start-up of a state-of-the-art fusion power plant. To gain access to the power plant control room, the student must complete several education modules, including topics on building an atom, fusion reactions, charged particle motion in electric and magnetic fields, and building a power plant. Review questions, a fusion video, library material and glossary provide additional resources. In the control room the student must start-up a complex, dynamic fusion power plant. The simulation model contains primary elements of a tokamak based device, including a magnetic shaper capable of producing limited and diverted elongated plasmas. A zero dimensional plasma model based on ITER scaling and containing rate based conservation equations provides dynamic feedback through major control parameters such as toroidal field, fueling rate and heating. The game is available for use on PC and Mac. computers. Copies will be available at the conference.

Modelling the power conversion unit of a generic nuclear fusion plant, with a dual coolant blanket and a supercritical CO2 power cycle, by means of RELAP5-3D

Energy Technology Data Exchange (ETDEWEB)

Batet, L.

2015-07-01

In the framework of the Spanish fusion program TECNO-FUS, a dual coolant blanket design was proposed for DEMO. A generic power conversion system (supercritical recompression CO2 cycle) based on this proposal has been simulated using RELAP5-3D, a multipurpose system thermal-hydraulic code developed by the Idaho National Laboratory (USA). The code allows the dynamic simulation of thermal-hydraulic systems, including the control features. A model has been set up by assembling the available RELAP5-3D components: pipe, branch, pump, compressor, turbine, etc. Thermal fluxes between fluids in heat exchangers are simulated by means of heat structures, which are used as well to simulate the heating from plasma. A number of control features have been designed for the simulated plant, and their parameters have been adjusted. The code is then able to simulate robustly the dynamics of the system with a few boundary conditions. This paper exemplifies the usefulness of the code and model to understand the behavior of the plant and to perform sensitivity analyses of the control parameters or other design features. (Author)

Status of inertial fusion and prospects for practical power plants

International Nuclear Information System (INIS)

Blink, J.A.; Monsler, M.J.

1982-01-01

We have produced a series of reactor designs to meet the variety of driver-target combinations that could possibly result from the inertial-confinement fusion program. In this paper we discuss four reactor designs, the goals of which are low cost; a low probability of risk to the public, the plant employees, and the utility investment; and a minimal environmental impact under normal plant operation. HYLIFE is a low pulse rate, lithium-cooled reactor. Pulse*Star and Cascade are high pulse rate reactors. In Pulse*Star, fusion energy is absorbed in the PbLi pool; in Cascade it is absorbed by Li 2 O particles. Sunburst, a very low pulse rate, lithium-cooled reactor, directly converts over 40% of the fusion energy to electricity using a pulsed magnetic field

Fusion tritium program in Japan

International Nuclear Information System (INIS)

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

1988-01-01

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

IFMIF suitability for evaluation of fusion functional materials

International Nuclear Information System (INIS)

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

2011-01-01

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

Overview of fusion nuclear technology in the US

International Nuclear Information System (INIS)

Morley, N.B.; Abdou, M.A.; Anderson, M.; Calderoni, P.; Kurtz, R.J.; Nygren, R.; Raffray, R.; Sawan, M.; Sharpe, P.; Smolentsev, S.; Willms, S.; Ying, A.Y.

2006-01-01

Fusion nuclear technology (FNT) research in the United States encompasses many activities and requires expertise and capabilities in many different disciplines. The US Enabling Technology program is divided into several task areas, with aspects of magnet fusion energy (MFE) fusion nuclear technology being addressed mainly in the Plasma Chamber, Neutronics, Safety, Materials, Tritium and Plasma Facing Component Programs. These various programs work together to address key FNT topics, including support for the ITER basic machine and the ITER Test Blanket Module, support for domestic plasma experiments, and development of DEMO relevant material and technological systems for blankets, shields, and plasma facing components. In addition, two inertial fusion energy (IFE) research programs conducting FNT-related research for IFE are also described. While it is difficult to describe all these activities in adequate detail, this paper gives an overview of critical FNT activities

PSO-Ensemble Demo Application

DEFF Research Database (Denmark)

2004-01-01

Within the framework of the PSO-Ensemble project (FU2101) a demo application has been created. The application use ECMWF ensemble forecasts. Two instances of the application are running; one for Nysted Offshore and one for the total production (except Horns Rev) in the Eltra area. The output...

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 1

International Nuclear Information System (INIS)

Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

1991-10-01

The BOT (Breeder Outside Tube) Helium Cooled Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test plankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

Status report. KfK contribution to the development of DEMO-relevant test blankets for NET/ITER. Pt. 2: BOT helium cooled solid breeder blanket. Vol. 2

International Nuclear Information System (INIS)

Dalle Donne, M.; Boccaccini, L.V.; Bojarsky, E.; Deckers, H.; Dienst, W.; Doerr, L.; Fischer, U.; Giese, H.; Guenther, E.; Haefner, H.E.; Hofmann, P.; Kappler, F.; Knitter, R.; Kuechle, M.; Moellendorf, U. von; Norajitra, P.; Penzhorn, R.D.; Reimann, G.; Reiser, H.; Schulz, B.; Schumacher, G.; Schwenk-Ferrero, A.; Sordon, G.; Tsukiyama, T.; Wedemeyer, H.; Weimar, P.; Werle, H.; Wiegner, E.; Zimmermann, H.

1991-10-01

The BOT (Breeder Outside Tube) Helium Cooled Solid Breeder Blanket for a fusion Demo reactor and the status of the R and D program is presented. This is the KfK contribution to the European Program for the Demo relevant test blankets to be irradiated in NET/ITER. Volume 1 (KfK 4928) contains the summary, volume 2 (KfK 4929) a more detailed version of the report. In both volumes are described the reasons for the selected design, the reference blanket design for the Demo reactor, the design of the test blanket including the ancillary systems together with the present status of the relative R and D program in the fields of neutronic and thermohydraulic calculations, of the electromagnetic forces caused by disruptions, of the development and irradiation of the ceramic breeder material, of the tritium release and recovery, and of the technological investigations. An outlook is given on the required R and D program for the BOT Helium Cooled Solid Breeder Blanket prior to tests in NET/ITER and the proposed test program in NET/ITER. (orig.) [de

Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Hartmann, Tobias

2013-07-03

The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

International Nuclear Information System (INIS)

Hartmann, Tobias

2013-01-01

The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

Availability of fusion plants employing a Monte Carlo simulation computer code

International Nuclear Information System (INIS)

Musicki, Z.

1984-01-01

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

Personnel Safety for Future Magnetic Fusion Power Plants

Energy Technology Data Exchange (ETDEWEB)

Lee Cadwallader

2009-07-01

The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

Personnel Safety for Future Magnetic Fusion Power Plants

International Nuclear Information System (INIS)

Cadwallader, Lee

2009-01-01

The safety of personnel at existing fusion experiments is an important concern that requires diligence. Looking to the future, fusion experiments will continue to increase in power and operating time until steady state power plants are achieved; this causes increased concern for personnel safety. This paper addresses four important aspects of personnel safety in the present and extrapolates these aspects to future power plants. The four aspects are personnel exposure to ionizing radiation, chemicals, magnetic fields, and radiofrequency (RF) energy. Ionizing radiation safety is treated well for present and near-term experiments by the use of proven techniques from other nuclear endeavors. There is documentation that suggests decreasing the annual ionizing radiation exposure limits that have remained constant for several decades. Many chemicals are used in fusion research, for parts cleaning, as use as coolants, cooling water cleanliness control, lubrication, and other needs. In present fusion experiments, a typical chemical laboratory safety program, such as those instituted in most industrialized countries, is effective in protecting personnel from chemical exposures. As fusion facilities grow in complexity, the chemical safety program must transition from a laboratory scale to an industrial scale program that addresses chemical use in larger quantity. It is also noted that allowable chemical exposure concentrations for workers have decreased over time and, in some cases, now pose more stringent exposure limits than those for ionizing radiation. Allowable chemical exposure concentrations have been the fastest changing occupational exposure values in the last thirty years. The trend of more restrictive chemical exposure regulations is expected to continue into the future. Other issues of safety importance are magnetic field exposure and RF energy exposure. Magnetic field exposure limits are consensus values adopted as best practices for worker safety; a typical

External costs of material recycling strategies for fusion power plants

International Nuclear Information System (INIS)

Hallberg, B.; Aquilonius, K.; Lechon, Y.; Cabal, H.; Saez, R.M.; Schneider, T.; Lepicard, S.; Ward, D.; Hamacher, T.; Korhonen, R.

2003-01-01

This paper is based on studies performed within the framework of the project Socio-Economic Research on Fusion (SERF3). Several fusion power plant designs (SEAFP Models 1-6) were compared focusing on part of the plant's life cycle: environmental impact of recycling the materials. Recycling was considered for materials replaced during normal operation, as well as materials from decommissioning of the plant. Environmental impact was assessed and expressed as external cost normalised with the total electrical energy output during plant operation. The methodology used for this study has been developed by the Commission of the European Union within the frame of the ExternE project. External costs for recycling, normalised with the energy production during plant operation, are very low compared with those for other energy sources. Results indicate that a high degree of recycling is preferable, at least when considering external costs, because external costs of manufacturing of new materials and disposal costs are higher

Fusion energy

International Nuclear Information System (INIS)

Gross, R.A.

1984-01-01

This textbook covers the physics and technology upon which future fusion power reactors will be based. It reviews the history of fusion, reaction physics, plasma physics, heating, and confinement. Descriptions of commercial plants and design concepts are included. Topics covered include: fusion reactions and fuel resources; reaction rates; ignition, and confinement; basic plasma directory; Tokamak confinement physics; fusion technology; STARFIRE: A commercial Tokamak fusion power plant. MARS: A tandem-mirror fusion power plant; and other fusion reactor concepts

Data fusion and sensor management for nuclear power plant safety

Energy Technology Data Exchange (ETDEWEB)

Ciftcioglu, O [Istanbul Technical Univ., Istanbul (Turkey). Nuclear Power Dept.; Turkcan, E [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

1997-12-31

The paper describes the implementation of the data-sensor fusion and sensor management technology for accident management through simulated severe accident (SA) scenarios subjected to study. The organization of the present paper is as follows. As the data-sensor fusion and sensor management is an emerging technology which is not widely known, in Sec. 2, the definition and goals of data-sensor fusion and sensor management technology is described. In Sec. 3 fits, with reference to Kalman filtering as an information filter, statistical data-sensor fusion technology is described. This is followed by deterministic data-sensor fusion technology using gross plant state variables and neural networks (NN) and the implementation for severe accident management in NPPs. In Sec. 4, the sensor management technology is described. Finally, the performance of the data-sensor fusion technology for NPP safety is discussed. 12 refs, 6 figs.

Data fusion and sensor management for nuclear power plant safety

International Nuclear Information System (INIS)

Ciftcioglu, O.

1996-01-01

The paper describes the implementation of the data-sensor fusion and sensor management technology for accident management through simulated severe accident (SA) scenarios subjected to study. The organization of the present paper is as follows. As the data-sensor fusion and sensor management is an emerging technology which is not widely known, in Sec. 2, the definition and goals of data-sensor fusion and sensor management technology is described. In Sec. 3 fits, with reference to Kalman filtering as an information filter, statistical data-sensor fusion technology is described. This is followed by deterministic data-sensor fusion technology using gross plant state variables and neural networks (NN) and the implementation for severe accident management in NPPs. In Sec. 4, the sensor management technology is described. Finally, the performance of the data-sensor fusion technology for NPP safety is discussed. 12 refs, 6 figs

Design aspects of a multipurpose fusion power plant for desalination and agrochemical processes

International Nuclear Information System (INIS)

Sabri, Z.A.

1975-02-01

A description is given of the skeletal structure of a multipurpose fusion power plant, designed for desalination and agrochemicals production. The plant is a complex that comprises dual purpose power and desalination units, separation and processing units for recovery of soluble salts in the effluent of the desalination unit, mariculture units for production of algae for food and as food for shrimp and other fish species. The electrical power unit is a two-component fusion device that burns deuterium and helium-3 utilizing a fast fusion cycle

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-15

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

Scoping studies for NBI launch geometries on DEMO

Energy Technology Data Exchange (ETDEWEB)

Jenkins, I., E-mail: ian.jenkins@ukaea.uk; Challis, C.D.; Keeling, D.L.; Surrey, E.

2016-05-15

Highlights: • NBCD scans are done for beam energies of 1.5 MeV and 1.0 MeV in two DEMO scenarios. • NBCD scan profiles are fed into genetic algorithm to fit a target current profile. • The result gives location and power of sources to give best fit to target profile. • This method can help provide requirements for DEMO beamline geometry. - Abstract: Engineering and technical constraints on Neutral Beam Injection (NBI) in DEMO may determine the available beam energy and may also strongly impact the Neutral Beam Current Drive (NBCD) efficiency by restricting available beam tangential radii. These latter are determined by factors such as the inter-TF coil spacing, as well as the degree of required shielding. In order to illustrate how these factors may affect the contribution of NBCD on DEMO operating scenarios, scans of NBI tangency radii and elevation on two possible DEMO scenarios have been performed with two beam energies, 1.5 MeV and 1.0 MeV, in order to determine the most favourable options for NBCD efficiency. In addition, a method using a genetic algorithm has been used to seek optimised solutions of NBI source locations and powers to attempt to synthesize a target total plasma driven-current profile. It is found that certain beam trajectories may be proscribed by limitations on shinethrough onto the vessel wall. This may affect the ability of NBCD to extend the duration of a pulse in a scenario where it must complement the induced plasma current. Operating at the lower beam energy reduces the restrictions due to shinethrough and is attractive for technical reasons as it will required less development, but in the scenarios examined here this results in a spatial broadening of the NBCD profile, which may make it more challenging to achieve desired total driven-current profiles.

The long range DIII-D plan

International Nuclear Information System (INIS)

Simonen, T.C.

1993-02-01

The mission of the DIII-D tokamak research program is to provide data needed by ITER and to develop a conceptual physics blueprint for a commercially attractive electrical demonstration plant (DEMO) that would open a path to fusion power commercialization. The National Energy Strategy calls for the development of magnetic fusion as an energy option with operation of a DEMO by 2025. The DEMO will be based on nuclear technology demonstrated in ITER and the physics and engineering database established in magnetic fusion facilities during the next two decades. On the present path, based on extrapolation of current conventional operating modes, ITER is twice as large as Joint European Tokamak (JET), and DEMO, using the same logic, will be even larger. However, successful development of advanced tokamak operating modes could open the way for significantly improved confinement and stability, leading to a smaller, more commercially attractive DEMO, provided new diverter concepts are developed to handle the accompanying high divertor power density. A smaller and lower cost DEMO opens up the possibility that multiple nations, utilities, and industries could build DEMOs simultaneously and, therefore, more rapidly optimize the tokamak for commercialization. Results from experiments at DIII-D and other tokamaks indicate that plasma and divertor performance can be increased transiently beyond the baseline conceptual design of ITER. A simultaneous long pulse demonstration of such improved tokamak plasma and divertor operation for steady state would establish an advanced physics foundation for the tokamak physics experiment program, provide new operating options for ITER, and open a path to an attractive DEMO. The planned DIII-D program incorporates new theory and technology developments to extend the tokamak experimental physics database toward steady state. This research program will also continue to provide increased understanding in many areas of fusion science and technology

Design of a Multi-Tube Pd-Membrane Module for Tritium Recovery from He in DEMO

Directory of Open Access Journals (Sweden)

Marco Incelli

2016-10-01

Full Text Available Dense self-supported Pd-alloy membranes are used to selectively separate hydrogen and hydrogen isotopes. In particular, deuterium (D and tritium (T are currently identified as the main elements for the sustainability of the nuclear fusion reaction aimed at carbon free power generation. In the fusion nuclear reactors, a breeding blanket produces the tritium that is extracted and purified before being sent to the plasma chamber in order to sustain the fusion reaction. In this work, the application of Pd-alloy membranes has been tested for recovering tritium from a solid breeding blanket through a helium purge stream. Several simulations have been performed in order to optimize the design of a Pd-Ag multi-tube module in terms of geometry, operating parameters, and membrane module configuration (series vs. parallel. The results demonstrate that a pre-concentration stage before the Pd-membrane unit is mandatory because of the very low tritium concentration in the He which leaves the breeding blanket of the fusion reactor. The most suitable operating conditions could be reached by: (i increasing the hydrogen partial pressure in the lumen side and (ii decreasing the shell pressure. The preliminary design of a membrane unit has been carried out for the case of the DEMO fusion reactor: the optimized membrane module consists of an array of 182 Pd-Ag tubes of 500 mm length, 10 mm diameter, and 0.100 mm wall thickness (total active area of 2.85 m2.

Optimization of the first wall for the DEMO water cooled lithium lead blanket

Energy Technology Data Exchange (ETDEWEB)

Aubert, Julien, E-mail: julien.aubert@cea.fr [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Aiello, Giacomo [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Bachmann, Christian [EFDA, Boltzmannstraße 2, 85748 Garching (Germany); Di Maio, Pietro Alessandro [Università di Palermo, Viale delle Scienze, 90128 Palermo (Italy); Giammusso, Rosario [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy); Li Puma, Antonella; Morin, Alexandre [CEA Saclay, F-91191 Gif-Sur-Yvette (France); Tincani, Amelia [ENEA C.R. Brasimone, 40032 Camugnano, Bologna (Italy)

2015-10-15

Highlights: • This paper presents the optimization of the first wall of the water cooled lithium lead DEMO blanket with pressurized water reactor condition and circular channels in order to find the best geometry that can allow the maximum heat flux considering design criteria since an estimate of the engineering limit of the first wall heat load capacity is an essential input for the decision to implement limiters in DEMO. • An optimization study was carried out for the flat first wall design of the DEMO Water-Cooled Lithium Lead considering thermal and mechanical constraint functions, assuming T{sub inlet}/T{sub outlet} equal to 285 °C/325 °C, based on geometric design parameters. • It became clear that through the optimization the advantages of a waved First Wall are diminished. • The analysis shows that the maximum heat load could achieve 2.53 MW m{sup −2}, but considering assumptions such as a coolant velocity ≤8 m/s, pipe diameter ≥5 mm and a total first wall thickness ≤22 mm, heat flux is limited to 1.57 MW m{sup −2}. - Abstract: The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analysis equal to 1.0 MW m{sup −2} with respect to the Eurofer temperature limit. An optimization study was then carried out for a flat FW design considering thermal and mechanical constraints assuming inlet and outlet

Environmental release targets for fusion power plants

International Nuclear Information System (INIS)

Gulden, W.; Raskob, W.

2005-01-01

Within the European fission community, so called European Utility Requirements were developed to define common targets, criteria and evaluation methods for, amongst others, safety, environmental protection and public health with respect to future nuclear fission power plant development. In the case of severe accidents, the objective is to restrict the radiological consequences to the vicinity of the plant, i.e., to avoid early and late countermeasures such as evacuation or relocation of the population, and to restrict food banning to small areas and the first year after the accident. Within the European Fusion Technology Programme, a methodology is being developed in compliance with these European Utility Requirements, to define design requirements for future fusion reactors. First results are presented. Concerning food banning, calculations revealed extremely conservative values for tritium in EU regulations and recommendations. This does not affect assessments for fission reactors, but is an overestimation of the tritium dose impact from ingestion. Therefore, in compliance with scientific justification, considerably higher maximum permissible activity levels for tritium should be considered

AC loss, interstrand resistance and mechanical properties of prototype EU DEMO TF conductors up to 30 000 load cycles

Science.gov (United States)

Yagotintsev, K.; Nijhuis, A.

2018-07-01

Two prototype Nb3Sn cable-in-conduit conductors conductors were designed and manufactured for the toroidal field (TF) magnet system of the envisaged European DEMO fusion reactor. The AC loss, contact resistance and mechanical properties of two sample conductors were tested in the Twente Cryogenic Cable Press under cyclic load up to 30 000 cycles. Though both conductors were designed to operate at 82 kA in a background magnetic field of 13.6 T, they reflect different approaches with respect to the magnet winding pack assembly. The first approach is based on react and wind technology while the second is the more common wind and react technology. Each conductor was tested first for AC loss in virgin condition without handling. The impact of Lorentz load during magnet operation was simulated using the cable press. In the press each conductor specimen was subjected to transverse cyclic load up to 30 000 cycles in liquid helium bath at 4.2 K. Here a summary of results for AC loss, contact resistance, conductor deformation, mechanical heat production and conductor stiffness evolution during cycling of the load is presented. Both conductors showed similar mechanical behaviour but quite different AC loss. In comparison with previously tested ITER TF conductors, both DEMO TF conductors possess very low contact resistance resulting in high coupling loss. At the same time, load cycling has limited impact on properties of DEMO TF conductors in comparison with ITER TF conductors.

Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

International Nuclear Information System (INIS)

Daigo Tsuru; Mikio Enoeda; Masato Akiba

2006-01-01

Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

Impurity accumulation and performance of ITER and DEMO plasmas in the presence of transport barriers

International Nuclear Information System (INIS)

Chatthong, B; Promping, J; Onjun, T

2017-01-01

In this work, the impurity accumulations and their performance in the presence of both ITB and ETB in ITER and DEMO plasmas are investigated using a BALDUR integrated predictive modelling code. In these simulations, a combination of a neoclassical transport model NCLASS and an anomalous transport model Mixed Bohm/gyro-Bohm is used. The boundary condition is described at the top of the pedestal, which is calculated theoretically based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The toroidal flow is calculated based on the NTV (neoclassical toroidal viscosity) toroidal velocity model. The time evolution of plasma temperature and density profiles of ITER and DEMO (Korean K-DEMO and Japanese DEMO models A, B and C) plasmas are simulated in H -mode scenario with and without ITB formation. It is found that Japanese DEMO model C yields highest plasma temperature; while Korean DEMO yields the best plasma performance among those designs considered. Impurity accumulation is found to be highest in Japanese DEMO model B. (paper)

Long-term ETR/INTOR magnet testing in support of the demonstration fusion reactor

International Nuclear Information System (INIS)

Herring, J.S.; Shah, V.N.; Rouhani, S.Z.

1983-01-01

This study considers ways that the proposed Engineering Test Reactor (ETR), or the proposed International Tokamak Reactor (INTOR), can be used for magnet performance tests that would be useful for the design and operation of the Demonstration Tokamak Power Plant (DEMO). Such testing must not interfere with the main function of the ETR/INTOR as an integrated fusion reactor. A performance test plan for the ETR/INTOR magnets is proposed and appropriate tests on the magnets is proposed and appropriate tests on the magnets for each phase of the ETR/INTOR operation are described. The suggested tests would verify design requirements and monitor long-term changes due to radiation. This paper also summarizes the design and operational performance of existing superconducting magnets and identifies the known failures and their predominant causes

Scenarios for multi-unit inertial fusion energy plants producing hydrogen fuel

International Nuclear Information System (INIS)

Logan, B.G.

1993-12-01

This work describes: (a) the motivation for considering fusion in general, and Inertial Fusion Energy (IFE) in particular, to produce hydrogen fuel powering low-emission vehicles; (b) the general requirements for any fusion electric plant to produce hydrogen by water electrolysis at costs competitive with present consumer gasoline fuel costs per passenger mile, for advanced car architectures meeting President Clinton's 80 mpg advanced car goal, and (c) a comparative economic analysis for the potential cost of electricity (CoE) and corresponding cost of hydrogen (CoH) from a variety of multi-unit IFE plants with one to eight target chambers sharing a common driver and target fab facility. Cases with either heavy-ion or diode-pumped, solid-state laser drivers are considered, with ''conventional'' indirect drive target gains versus ''advanced, e.g. Fast Ignitor'' direct drive gain assumptions, and with conventional steam balance-of-plant (BoP) versus advanced MHD plus steam combined cycle BoP, to contrast the potential economics under ''conventional'' and ''advanced'' IFE assumptions, respectively

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

The role of risk management in the design of diagnostics for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Ingesson, L. C. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Collaboration: F4E Diagnostic Project Team

2014-08-21

A project-oriented approach is beneficial for the selection and design of viable diagnostics for fusion reactors because of the associated complex physical and organizational environment. The project-oriented approach includes rigorous risk management. The nature and impact of risks related to technical, organizational and commercial aspects in relation to the development of ITER diagnostics under EU responsibility are analyzed. The majority of risks are related to organizational aspects and technical feasibility issues. The experience with ITER is extrapolated to DEMO and beyond. It should not be taken for granted that technical solutions will be found, while a risk analysis of various diagnostic techniques with quantitative assessments undertaken early in the design of DEMO would be beneficial.

Economic comparison of fusion power plant designs

International Nuclear Information System (INIS)

O'Neill, J.E.

1986-01-01

Over the past 10 yr, a number of studies have been developed for fusion power plants of various types (tokamaks, mirrors, etc.) complete with figures of merit such as cost estimates and estimates of the cost of generating electricity (COE). Each of these designs involves unresolved physics and engineering problems which, it is assumed, will eventually be worked out. Because of such uncertainties the figures of merit associated with such designs are not to be compared as absolute measures of worth but as relative indicators of progress within a given concept type. As part of Grumman's involvement in fusion energy development, an effort has been undertaken to compare economic indicators from the referenced studies in order to determine the cost trend in recent reactor design activities

Status and strategy of fusion materials development in China

International Nuclear Information System (INIS)

Huang, Q.Y.; Wu, Y.C.; Li, J.G.; Wan, F.R.; Chen, J.L.; Luo, G.N.; Liu, X.; Chen, J.M.; Xu, Z.Y.; Zhou, X.G.; Ju, X.; Shan, Y.Y.; Yu, J.N.; Zhu, S.Y.; Zhang, P.Y.; Yang, J.F.; Chen, X.J.; Dong, S.M.

2009-01-01

The liquid metal and solid ceramic pebble breeder blankets have become the most promising blankets for ITER-TBMs or DEMO reactors in China and the world due to their potential advantages. In recent years the corresponding research work on fusion reactor materials mainly focuses on structural materials, plasma facing materials and the functional materials for the blanket such as breeder, coating and flow channel insert etc. for the successful application of fusion energy in the near future. The R and D on those materials in the two kinds of blankets is being carried out widely in China, including fabrication and manufacturing techniques, physical/mechanical properties assessment before and after irradiation, joining techniques for structural materials, compatibility evaluation, and the development and verification of the criteria for fusion material designs. The progress on main R and D activities of fusion reactor materials in China is introduced and prospected in the paper.

Preconceptual design of hyfire. A fusion driven high temperature electrolysis plant

International Nuclear Information System (INIS)

Varljen, T.C.; Chi, J.W.H.; Karbowski, J.S.

1983-01-01

Brookhaven National Laboratory has been engaged in a scoping study to investigate the potential merits of coupling a fusion reactor with a high temperature blanket to a high temperature electrolysis (HTE) process to produce hydrogen and oxygen. Westinghouse is assisting this study in the areas of systems design integration, plasma engineering, balance of plant design and electrolyzer technology. The aim of the work done in the past year has been to focus on a reference design point for the plant, which has been designated HYFIRE. In prior work, the STARFIRE commercial tokamak fusion reactor was directly used as the fusion driver. This report describes a new design obtained by scaling the basic STARFIRE design to permit the achievement of a blanket power of 6000 MWt. The high temperature blanket design employs a thermally insulated refractory oxide region which provides high temperature (>1000 deg. C) steam at moderate pressures to high temperature electrolysis units. The electrolysis process selected is based on the high temperature, solid electrolyte fuel cell technology developed by Westinghouse. An initial process design and plant layout has been completed; component cost and plant economics studies are now underway to develop estimates of hydrogen production costs and to determine the sensitivity of this cost to changes in major design parameters. (author)

Conceptual design of a laser fusion power plant

International Nuclear Information System (INIS)

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

1977-01-01

A conceptual design of a laser fusion power plant is extensively discussed. Recent advances in high gain targets are exploited in the design. A smaller blanket structure is made possible by use of a thick falling region of liquid lithium for a first wall. Major design features of the plant, reactor, and laser systems are described. A parametric analysis of performance and cost vs. design parameters is presented to show feasible design points. A more definitive follow-on conceptual design study is planned

High current superconductors for DEMO

Energy Technology Data Exchange (ETDEWEB)

Bruzzone, Pierluigi, E-mail: pierluigi.bruzzone@psi.ch [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom – Confédération Suisse, CH-5232 Villigen PSI (Switzerland); Sedlak, Kamil; Stepanov, Boris [Ecole Polytechnique Fédérale de Lausanne (EPFL), Centre de Recherches en Physique des Plasmas (CRPP), Association Euratom – Confédération Suisse, CH-5232 Villigen PSI (Switzerland)

2013-10-15

Highlights: ► Definition of requirement for TF coil based on the input of system code. ► A TF coil and conductor design for the European DEMO project. ► Use of React and Wind method opposite to Wind and React with related advantages. ► Hybridization of winding pack, Nb/Nb{sub 3}Sn, by graded layer winding. -- Abstract: In the assumption that DEMO will be an inductively driven tokamak, the number of load cycles will be in the range of several hundred thousands. The requirements for a new generation of Nb{sub 3}Sn based high current conductors for DEMO are drafted starting from the output of system code PROCESS. The key objectives include the stability of the DC performance over the lifetime of the machine and the effective use of the Nb{sub 3}Sn strand properties, for cost and reliability reasons. A preliminary layout of the winding pack and conductors for the toroidal field magnets is presented. To suppress the mechanism of reversible and irreversible degradation, i.e. to preserve in the cabled conductor the high critical current density of the strand, the thermal strain must be insignificant and no space for micro-bending under transverse load must be left in the strand bundle. The “react-and-wind” method is preferred here, with a graded, layer wound magnet, containing both Nb{sub 3}Sn and NbTi layers. The implications of the conductor choice on the coil design and technology are highlighted. A roadmap is sketched for the development of a full size prototype conductor sample and demonstration of the key technologies.

Scoping the parameter space for demo and the engineering test

International Nuclear Information System (INIS)

Meier, W R.

1999-01-01

In our IFE development plan, we have set a goal of building an Engineering Test Facility (ETF) for a total cost of $2B and a Demo for $3B. In Mike Campbell s presentation at Madison, we included a viewgraph with an example Demo that had 80 to 250 MWe of net power and showed a plausible argument that it could cost less than $3B. In this memo, I examine the design space for the Demo and then briefly for the ETF. Instead of attempting to estimate the costs of the drivers, I pose the question in a way to define R ampersand D goals: As a function of key design and performance parameters, how much can the driver cost if the total facility cost is limited to the specified goal? The design parameters examined for the Demo included target gain, driver energy, driver efficiency, and net power output. For the ETF; the design parameters are target gain, driver energy, and target yield. The resulting graphs of allowable driver cost determine the goals that the driver R ampersand D programs must seek to meet

Comparison of environmental impact of waste disposal from fusion, fission and coal-fired power plants

Energy Technology Data Exchange (ETDEWEB)

Frey, Bruno [Fichtner GmbH und Co. KG, Stuttgart (Germany)

2011-08-15

The radiotoxic hazard of waste from fusion power plants has been compared with that of fission power and radioactive trace elements in coal ash within some research programs such as SEAFP and SEIF. Within another program, in 2005 a Power Plant Conceptual Study (PPCS) has been finalized investigating 4 fusion power plant models A to D. In this paper, the radiotoxicity of model B is compared with a fission power plant, concentrating on the production of wastes. The hazard of the respective masses of enriched uranium before use in a fission power plant and coal ash of a power plant generating the same amount of electricity are used as benchmarks. It is evident that the development of ingestion and inhalation hazard of the PPCS model B is different from the results of earlier studies because of different assumptions on material impurities and other constraints. An important aspect is the presence of actinides in fusion power plant waste. (orig.)

Beryllium for fusion application - recent results

International Nuclear Information System (INIS)

Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

2002-01-01

The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described

Beryllium for fusion application - recent results

Science.gov (United States)

Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

2002-12-01

The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described.

Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

Energy Technology Data Exchange (ETDEWEB)

Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Gironimo, G. Di, E-mail: giuseppe.digironimo@unina.it [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Esposito, G. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Mäkinen, H. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Miccichè, G. [ENEA Brasimone, I:40032 Camugnano (Italy); Mozzillo, R. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy)

2016-11-15

Highlights: • Concept Studies in Divertor Remote Handling. • Prioritization of concept alternatives. • Comparison and evaluation of product alternatives using AHP. - Abstract: The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

Multicriteria selection in concept design of a divertor remote maintenance port in the EU DEMO reactor using an AHP participative approach

International Nuclear Information System (INIS)

Carfora, D.; Gironimo, G. Di; Esposito, G.; Huhtala, K.; Määttä, T.; Mäkinen, H.; Miccichè, G.; Mozzillo, R.

2016-01-01

Highlights: • Concept Studies in Divertor Remote Handling. • Prioritization of concept alternatives. • Comparison and evaluation of product alternatives using AHP. - Abstract: The work behind this paper took place in the Eurofusion remote maintenance system project (WPRM) for the EU Demonstration Fusion Power Reactor (DEMO). Following ITER, the aim of DEMO is to demonstrate the capability of generating several hundreds of MW of net electricity by 2050. The main objective of this paper was the study of the most efficient design of the maintenance port for replacing the divertor cassettes in a Remote Handling (RH) point of view. In DEMO overall design, one important consideration is the availability and short down time operations. The inclination of the divertor port has a very important impact on all the RH tasks such as the design of the divertor mover, the divertor locking systems and the end effectors. The current reference scenario of the EU DEMO foresees a 45° inclined port for the remote maintenance (RM) of the divertor in the lower part of the reactor. Nevertheless, in the optic of the systems engineering (SE) approach, in early concept design phase, all possible configurations shall be taken into account. Even the solutions which seem not feasible at all need to be investigated, because they could lead to new and innovative engineering proposals. The different solutions were compared using an approach based on the Analytic Hierarchy Process (AHP). The technique is a multi-criteria decision making approach in which the factors that are important in making a decision are arranged in a hierarchic structure. The results of these studies show how the application of the AHP improved and focused the selection on the concept which is closer to the requirements arose from technical meetings with the experts of the RH field.

Design windows of laser fusion power plants and conceptual design of laser-diode pumped slab laser

International Nuclear Information System (INIS)

Kozaki, Y.; Eguchi, T.; Izawa, Y.

1999-01-01

An analysis of the design space available to laser fusion power plants has been carried out, in terms of design key parameters such as target gain, laser energy and laser repetition rate, the number of fusion react ion chambers, and plant size. The design windows of economically attractive laser fusion plants is identified with the constraints of key design parameters and the cost conditions. Especially, for achieving high repetition rate lasers, we have proposed and designed a diode-pumped solid-state laser driver which consists of water-cooled zig-zag path slab amplifiers. (author)

Economic and environmental performance of future fusion plants in comparison

International Nuclear Information System (INIS)

Hamacher, T.; Saez, R.M.; Lako, P.

2001-01-01

If the good performance of fusion as technology with no CO 2 emission during normal operation and rather low external costs, reflecting the advantageous environmental and safety characteristics, are considered in future energy regulations, fusion can win considerable market shares in future electricity markets. The economic performance was elaborated for Western Europe for the time period till 2100. The software tool MARKAL widely used in energy research was used to simulate and optimise the development of the Western European energy system. Two different scenarios were considered, the main difference was the interest rate for investments. Stringent CO 2 -emission strategies lead to considerable market shares for fusion. As a comprehensive indicator of the environmental and safety performance of fusion plants the external costs following the ExternE method was used. External costs of fusion are rather low, much below the cost of electricity, and are in the same range as photovoltaics and wind energy. (author)

Evaluation of divertor conceptual designs for a fusion power plant

International Nuclear Information System (INIS)

Ferrari, M.; Giancarli, L.; Kleefeldt, K.; Nardi, C.; Roedig, M.; Reimann, J.; Salavy, J.F.

2001-01-01

In the frame of the preliminary study of plants suitable for the energy production from the fusion power, particular emphasis has been given on the divertor studies. Since a significant percentage of the power generated from the fusion process is absorbed in the divertor, the thermal efficiency of the power conversion cycle requires a high coolant outlet temperature of the divertor, leading to solutions that are different from those adopted for the present experimental fusion plants. Therefore, copper alloys having extremely high thermal conductivity, cannot be used as structural material for this kind of devices. The most suitable coolants to be used in the divertor are water, helium and liquid metals. A conceptual design study has been developed for each of these three fluids, with the aim to evaluate the maximum allowable thermal flux at the divertor target plate and the R and D requirements for each solution. While a water-cooled divertor can be designed with a limited R and D effort, the development of helium or liquid metal cooled divertors requires a more engaging R and D program

Fusion research at Culham site

International Nuclear Information System (INIS)

Tolonen, P.; Toppila, T.

1998-01-01

One of the many targets on the Finnish Nuclear Society (ATS) excursion to England was the Culham fusion research site. The site has divided into two parts. One of them is UKAEA Fusion with small scale fusion reactors and 200 employees. UKAEA has 3 fusion reactors at Culham site. One of is the START (Small Tight Aspect Ratio Tokamak) which was operational since 1991 but is today already out of operation. UKAEA has been operating a JET-like tokamak fusion reactor COMPASS-D since 1989. The latest of three reactors is MAST (Mega Amp Spherical Tokamak), which is still under construction. The first plasma will take place in the end of 1998. Another part of Culham site is JET (Joint European Torus), an all-European fusion undertaking with 350 employees. 150 of them are from various European countries and the rest 200 are employed by UKAEA. JET is the biggest fusion reactor ever and it represents the latest step in world wide fusion programme. In October 1997 JET achieved a world record in fusion power and energy. JET produced 16,1 MW power for 1 s and totally 21,7 MJ energy. This is the closest attempt to achieve break-even conditions. The next step in world wide fusion programme will be international ITER-reactor. This undertaking has some financial problems, since United States has taken distance to magnetic fusion research and moved closer to inertial fusion with funding of US Department of Defence. The planned reactor, however, is physically twice as big as JET. The step after this phase will be DEMO, which is purposed to produce fusion energy. According to our hosts in Culham this phase is 40 years ahead. (author)

Effect on the Tritium Breeding Ratio due to a distributed ICRF antenna in a DEMO reactor

International Nuclear Information System (INIS)

Garcia, A.; Noterdaeme, J.-M.; Fischer, U.; Dies, J.

2016-01-01

This thesis reports results of MCNP-5 calculations, with the nuclear data library FENDL-2.1, to assess the effect on the Tritium Breeding Ratio (TBR) due to a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna integrated in the blanket of a DEMO fusion power reactor. A preliminary design of the antenna with a reference configuration of the DEMO reactor was used together with a parametric analysis for different parameters that strongly affect the TBR. These are the type of breeding blanket (Helium Cooled Pebble Bed, Helium Cooled Lithium Lead and Water Cooled Lithium Lead), the covering ratio of the straps of the antenna (the ratio between the surface of all the straps and the projected surface of the antenna slot: 0.49, 0.72 and 0.94), the antenna radial thickness (20 cm and 40 cm), the thickness of the straps (2 cm, 4 cm and a double layer of 0.2 cm plus 2.5 cm with the composition of the First Wall), and finally the poloidal position of the antenna (0°, which is the equatorial port, 40° and 90°, which is the upper port). For an antenna with a full toroidal circumference of 360°, located poloidaly at 40° with a poloidal extension of 1 m and a total First Wall surface of 67 m"2, the reduction of the TBR is −0.35% for a HCPB blanket concept, −0.53% for a HCLL blanket concept and −0.51% for a WCLL blanket concept. In all cases covered by the parametric analysis, the loss of TBR remains below 0.61%. Such a distributed ICRF antenna has thus only a marginal effect on the TBR for a DEMO reactor.

Effect on the Tritium Breeding Ratio due to a distributed ICRF antenna in a DEMO reactor

Energy Technology Data Exchange (ETDEWEB)

Garcia, A., E-mail: albert.garcia.hp@gmail.com [Max-Planck-Institut für Plasmaphysik (IPP), Garching (Germany); Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Polytechnic University of Catalonia (UPC), Barcelona (Spain); Department of Applied Physics, Ghent University, Ghent (Belgium); Noterdaeme, J.-M. [Max-Planck-Institut für Plasmaphysik (IPP), Garching (Germany); Department of Applied Physics, Ghent University, Ghent (Belgium); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Dies, J. [Polytechnic University of Catalonia (UPC), Barcelona (Spain)

2016-11-15

This thesis reports results of MCNP-5 calculations, with the nuclear data library FENDL-2.1, to assess the effect on the Tritium Breeding Ratio (TBR) due to a distributed Ion Cyclotron Range of Frequencies (ICRF) antenna integrated in the blanket of a DEMO fusion power reactor. A preliminary design of the antenna with a reference configuration of the DEMO reactor was used together with a parametric analysis for different parameters that strongly affect the TBR. These are the type of breeding blanket (Helium Cooled Pebble Bed, Helium Cooled Lithium Lead and Water Cooled Lithium Lead), the covering ratio of the straps of the antenna (the ratio between the surface of all the straps and the projected surface of the antenna slot: 0.49, 0.72 and 0.94), the antenna radial thickness (20 cm and 40 cm), the thickness of the straps (2 cm, 4 cm and a double layer of 0.2 cm plus 2.5 cm with the composition of the First Wall), and finally the poloidal position of the antenna (0°, which is the equatorial port, 40° and 90°, which is the upper port). For an antenna with a full toroidal circumference of 360°, located poloidaly at 40° with a poloidal extension of 1 m and a total First Wall surface of 67 m{sup 2}, the reduction of the TBR is −0.35% for a HCPB blanket concept, −0.53% for a HCLL blanket concept and −0.51% for a WCLL blanket concept. In all cases covered by the parametric analysis, the loss of TBR remains below 0.61%. Such a distributed ICRF antenna has thus only a marginal effect on the TBR for a DEMO reactor.

Issues and strategies for DEMO in-vessel component integration

International Nuclear Information System (INIS)

Bachmann, C.; Arbeiter, F.; Boccaccini, L.V.; Coleman, M.; Federici, G.; Fischer, U.; Kemp, R.; Maviglia, F.; Mazzone, G.; Pereslavtsev, P.; Roccella, R.; Taylor, N.; Villari, R.; Villone, F.; Wenninger, R.; You, J.-H.

2016-01-01

In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

Issues and strategies for DEMO in-vessel component integration

Energy Technology Data Exchange (ETDEWEB)

Bachmann, C., E-mail: christian.bachmann@euro-fusion.org [EUROfusion PMU, Garching (Germany); Arbeiter, F.; Boccaccini, L.V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Coleman, M.; Federici, G. [EUROfusion PMU, Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kemp, R. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Maviglia, F. [EUROfusion PMU, Garching (Germany); Mazzone, G. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Roccella, R. [ITER Organization, St. Paul Lez Durance (France); Taylor, N. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Villari, R. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Villone, F. [ENEA-CREATE Association, DIEI, Università di Cassino e del Lazio Meridiona (Italy); Wenninger, R. [EUROfusion PMU, Garching (Germany); You, J.-H. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany)

2016-11-15

In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

Časopis Demos (Internationale Ethnographische und Folkloristische Informationen) v novém tisíciletí

Czech Academy of Sciences Publication Activity Database

Woitsch, Jiří

2003-01-01

Roč. 6, - (2003), s. 74-78 ISSN 1210-1109 Institutional research plan: CEZ:AV0Z9058907 Keywords : journal Demos * history of the journal Demos * new developments in publishing of the journal Demos Subject RIV: AC - Archeology, Anthropology, Ethnology

Fusion: from sacred cow to white elephant?

International Nuclear Information System (INIS)

Wooldridge, J.

1994-01-01

Controlled thermonuclear fusion has the potential to supply lots of relatively cheap power relatively cheaply. It is also renewable and has public support. Because of this potential, fusion has been able to attract huge research funds. The four main research programmes, in Europe, USA, Japan and Russia, include cooperation on the International Thermonuclear Experimental Reactor, ITER. The siting of this reactor will be decided in 1998 and it is due to start operation in 2010. It should lead to a demonstration reactor, DEMO, after which a prototype commercial reactor is envisaged for 2030-2050. But this is too far away to solve some of the immediate energy problems such as carbon dioxide emissions and global warming. So even if the technical problems are solved, fusion may not be the wonder energy source when it finally arrives; the trend is away from centralised, high cost, high output generation. Fusion research has taken interest and money away from other alternatives such as tidal energy, fuel cells and photovoltaic cells. Photovoltaics in particular look more feasible than fusion and could be in place far sooner, but lack the funding for research. (UK)

Limitation of fusion power plant installation on future power grids under the effect of renewable and nuclear power sources

Energy Technology Data Exchange (ETDEWEB)

Takeda, Shutaro, E-mail: takeda.shutarou.55r@st.kyoto-u.ac.jp [Graduate School of Advanced Integrated Studies in Human Survivability, Kyoto University, Kyoto, Kyoto (Japan); Sakurai, Shigeki [Graduate School of Advanced Integrated Studies in Human Survivability, Kyoto University, Kyoto, Kyoto (Japan); Yamamoto, Yasushi [Faculty of Engineering Science, Kansai University, Suita, Osaka (Japan); Kasada, Ryuta; Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Uji, Kyoto (Japan)

2016-11-01

Graphical abstract: - Highlights: • Future power grids would be unstable due to renewable and nuclear power sources. • Output interruptions of fusion plant would cause disturbances to future grids. • Simulation results suggested they would create limitations in fusion installation. • A novel diagram was presented to illustrate this suggested limitation. - Abstract: Future power grids would be unstable because of the larger share of renewable and nuclear power sources. This instability might bring some additional difficulties to fusion plant installation. Therefore, the authors carried out a quantitative feasibility study from the aspect of grid stability through simulation. Results showed that the more renewable and nuclear sources are linked to a grid, the greater disturbance the grid experiences upon a sudden output interruption of a fusion power plant, e.g. plasma disruption. The frequency deviations surpassed 0.2 Hz on some grids, suggesting potential limitations of fusion plant installation on future grids. To clearly show the suggested limitations of fusion plant installations, a novel diagram was presented.

Tritium supply assessment for ITER and DEMOnstration power plant

International Nuclear Information System (INIS)

Ni, Muyi; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

2013-01-01

Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource

Tritium supply assessment for ITER and DEMOnstration power plant

Energy Technology Data Exchange (ETDEWEB)

Ni, Muyi, E-mail: muyi.ni@fds.org.cn; Wang, Yongliang; Yuan, Baoxin; Jiang, Jieqiong; Wu, Yican

2013-10-15

Highlights: • The tritium production rate in CANDU reactor was simulated and estimated. • Possible routes, including APT, CLWR and tritium production schemes of ADS, were evaluated in feasibility and economy. • The possible tritium consumption of ITER and initial supply for DEMO was assessed. • Result of supply and demand showed that after ITER retired in 2038, the tritium production in CANDU reactor might not be enough for a FDS-II scale DEMO reactor startup if without additional tritium resource. -- Abstract: The International Thermonuclear Experimental Reactor (ITER) and next generation DEMOnstration fusion reactor need amounts of tritium for test/initial startup and will consume kilograms tritium for operation per year. The available supply of tritium for fusion reactor is man-made sources. Now most of commercial tritium resource is extracted from moderator and coolant of CANada Deuterium Uranium (CANDU) type Heavy Water Reactor (HWR), in the Ontario Hydro Darlington facility of Canada and Wolsong facility of Korea. In this study, the tritium production rate in CANDU reactor was simulated and estimated. And other possible routes, including Accelerator Production of Tritium (APT), tritium production in Commercial Light Water Reactor (CLWR) and Accelerator Driven Subcritical system (ADS), were also evaluated in feasibility and economy. Based on the tritium requirement investigated according to ITER test schedule and startup inventory required for a FDS-II-scale DEMO calculated by TAS1.0, the assessment results showed that after ITER retired in 2038, the tritium inventory of CANDU reactor could not afford DEMO reactor startup without extra resource.

Conceptual design of a laser fusion power plant. Part I. An integrated facility

International Nuclear Information System (INIS)

1981-07-01

This study is a new preliminary conceptual design and economic analysis of an inertial confinement fusion (ICF) power plant performed by Bechtel under the direction of Lawrence Livermore National Laboratory (LLNL). The purpose of a new conceptual design is to examine alternatives to the LLNL HYLIFE power plant and to incorporate information from the recent liquid metal cooled power plant conceptual design study (CDS) into the reactor system and balance of plant design. A key issue in the design of a laser fusion power plant is the degree of symmetry in the illumination of the target that will be required for a proper burn. Because this matter is expected to remain unresolved for some time, another purpose of this study is to determine the effect of symmetry requirements on the total plant size, layout, and cost

European DEMO divertor target: Operational requirements and material-design interface

Directory of Open Access Journals (Sweden)

J.H. You

2016-12-01

Full Text Available Recently, an integrated program of conceptual design activities for the European DEMO reactor was launched in the framework of the EUROfusion Consortium, where reliable power handling capability was identified as one of the most critical scientific as well as technological challenges for a DEMO reactor. The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. The DEMO divertor target will have to withstand extreme thermal loads where the local peak heat flux is expected to reach up to 20 MW/m2 during slow transient events in DEMO. To assure sufficient heat removal capability of the divertor target against normal and transient operational scenarios under expected cumulative neutron dose of up to 13 dpa is one of the fundamental engineering challenges imposed on target design. To develop the design of the DEMO divertor and related technologies, an R&D work package ‘Divertor’ has been set up in this consortium. The subproject ‘Target Development’ is devoted to the development of the conceptual design and the core technologies of the plasma-facing target. Devising and implementing novel structural heat sink materials (e.g. W/Cu composites to advanced target design concepts is one of the major objectives of this subproject. In this paper, the underlying design requirements imposed by the envisaged power exhaust goal and the prominent material-design interface issues are discussed. In addition, the candidate design concepts being currently considered are presented together with the related material issues. Finally, the first results achieved so far are presented.

Divertor conceptual designs for a fusion power plant

International Nuclear Information System (INIS)

Norajitra, P.; Ihli, T.; Janeschitz, G.; Abdel-Khalik, S.; Mazul, I.; Malang, S.

2007-01-01

The development of a divertor concept for post-ITER fusion power plants is deemed to be an urgent task to meet the EU Fast Track scenario. Developing a divertor is particularly challenging due to the wide range of requirements to be met including the high incident peak heat flux, the blanket design with which the divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incident a particles, radiation effects on the properties of structural materials, and efficient recovery and conversion of the divertor thermal power (∝15% of the total fusion thermal power) by maximizing the coolant operating temperature while minimizing the pumping power. In the course of the EU PPCS, three near-term (A, B and AB) and two advanced power plant models (C, D) were investigated. Model A utilizes a water-cooled lead-lithium (WCLL) blanket and a water-cooled divertor with a peak heat flux of 15 MW/m 2 . Model B uses a He-cooled ceramics/beryllium pebble bed (HCPB) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model AB uses a He-cooled lithium-lead (HCLL) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model C is based on a dual-coolant (DC) blanket (lead/lithium self-cooled bulk and He-cooled structures) and a He-cooled divertor (10 MW/m 2 ). Model D employs a self-cooled lead/lithium (SCLL) blanket and lead-lithiumcooled divertor (5 MW/m 2 ). The values in parenthesis correspond to the maximum peak heat fluxes required. It can be noted that the helium-cooled divertor is used in most of the EU plant models; it has also been proposed for the US ARIES-CS reactor study. Since 2002, it has been investigated extensively in Europe under the PPCS with the goal of reaching a maximum heat flux of at least 10 MW/m2. Work has covered many areas including conceptual design, analysis, material and fabrication issues, and experiments. Generally, the helium-cooled divertor is considered to be a suitable solution for fusion power plants, as it

Comparison of material irradiation conditions for fusion, spallation, stripping and fission neutron sources

International Nuclear Information System (INIS)

Vladimirov, P.; Moeslang, A.

2004-01-01

Selection and development of materials capable of sustaining irradiation conditions expected for a future fusion power reactor remain a big challenge for material scientists. Design of other nuclear facilities either in support of the fusion materials testing program or for other scientific purposes presents a similar problem of irradiation resistant material development. The present study is devoted to an evaluation of the irradiation conditions for IFMIF, ESS, XADS, DEMO and typical fission reactors to provide a basis for comparison of the data obtained for different material investigation programs. The results obtained confirm that no facility, except IFMIF, could fit all user requirements imposed for a facility for simulation of the fusion irradiation conditions

Osiris and SOMBRERO inertial fusion power plant designs - summary, conclusions, and recommendations

International Nuclear Information System (INIS)

Meier, Wayne R.

1994-01-01

An 18 month study to evaluate the potential of inertial fusion energy (IFE) for electric power production has been completed. The primary objective of the study was to provide the US Department of Energy with an evaluation of the potential of inertial fusion for electric power production. The study included the conceptual design of two inertial fusion power plants. Osiris uses an induction linac heavy ion beam driver, and SOMBRERO uses a krypton fluoride laser driver. Conceptual designs were completed for the reactors, power conversion and plant facilities, and drivers. Environmental and safety aspects, technical issues, technology development needs, and economics of the final point designs were assessed and compared. This paper summarizes the results and conclusions of the conceptual designs and results of the assessment studies. We conclude that IFE has the potential of producing technically credible designs with environmental, safety, and economics characteristics that are just as attractive as magnetic fusion. Realizing this potential will require additional research and development on target physics, chamber design, target production and injection systems, and drivers. ((orig.))

Reel success creating demo reels and animation portfolios

CERN Document Server

Cabrera, Cheryl

2013-01-01

Are you an animator looking to get your foot in the door to the top studios?It's tough if you don't have a demo reel and portfolio that reflects your unique style and incredible talents.  The reception of that reel will make or break you; so it's no wonder that creating a demo reel can be such a daunting task.  Reel Success by Cheryl Cabrera can help.  This book guides you into putting the right content into your portfolio, how to cater to the right audience, and how to harness the power of social media and network effectively.  Accompanied by case studies of actual students

Dominant integration locus drives continuous diversification of plant immune receptors with exogenous domain fusions.

Science.gov (United States)

Bailey, Paul C; Schudoma, Christian; Jackson, William; Baggs, Erin; Dagdas, Gulay; Haerty, Wilfried; Moscou, Matthew; Krasileva, Ksenia V

2018-02-19

The plant immune system is innate and encoded in the germline. Using it efficiently, plants are capable of recognizing a diverse range of rapidly evolving pathogens. A recently described phenomenon shows that plant immune receptors are able to recognize pathogen effectors through the acquisition of exogenous protein domains from other plant genes. We show that plant immune receptors with integrated domains are distributed unevenly across their phylogeny in grasses. Using phylogenetic analysis, we uncover a major integration clade, whose members underwent repeated independent integration events producing diverse fusions. This clade is ancestral in grasses with members often found on syntenic chromosomes. Analyses of these fusion events reveals that homologous receptors can be fused to diverse domains. Furthermore, we discover a 43 amino acid long motif associated with this dominant integration clade which is located immediately upstream of the fusion site. Sequence analysis reveals that DNA transposition and/or ectopic recombination are the most likely mechanisms of formation for nucleotide binding leucine rich repeat proteins with integrated domains. The identification of this subclass of plant immune receptors that is naturally adapted to new domain integration will inform biotechnological approaches for generating synthetic receptors with novel pathogen "baits."

Needs and gaps in the development of aluminum-based corrosion and T-permeation barriers for DEMO blankets

Energy Technology Data Exchange (ETDEWEB)

Wulf, Sven-Erik, E-mail: sven-erik.wulf@kit.edu; Krauss, Wolfgang; Konys, Jürgen

2015-10-15

Highlights: • New processes for barriers based on electroplating introduced in the last years. • New processes ECA and ECX able to overcome former fabrication problems. • Scales by ECA showed long-term compatibility in flowing Pb–Li (>12,000 h). • Further fusion relevant characterization and optimization of scales is required. • Qualification of T-permeation properties is urgently needed. - Abstract: Low-activation-ferritic–martensitic (RAFM) steels are candidates for structural materials in different blanket designs foreseen for DEMO and partly for TBM's tested in ITER. In all designs the liquid breeder Pb–15.7Li is in direct contact with the structural material, and thus two major topics – corrosion and T-permeation – influence the reliable, safe and economical application of such combination of breeder and structural material. As bare RAFM steels exhibit high corrosion rates of up to 400 μm/h in flowing Pb–15.7Li, Al-based coatings made by different coating processes were developed during the last 15 years and showed promising results in protecting RAFM steels from corrosion and T-permeation reduction. Especially barriers made by HDA, and electroplating (ECA, ECX), proved their ability to protect Eurofer against corrosion in flowing Pb–15.7Li. However, available T-permeation data for coated RAFM steels are rare and partly ambiguous for these coatings. This paper summarizes the state-of-the-art of aluminum-based barrier development and points out gaps and needs in future scale characterization and T-permeation barrier development. Additionally, necessary qualification steps on the path toward a reliable fabrication route are presented that is required to produce aluminum-based corrosion and T-permeation barriers on RAFM steels for blanket applications in future fusion reactors like DEMO.

Relevance of NET first wall concept for DEMO DN

International Nuclear Information System (INIS)

Kiltie, J.S.

1987-01-01

Design studies for the Next European Torus (NET) have produced a design concept for the first wall. This concept features poloidal water cooling, double contained in a welded steel structure which is protected by radiatively cooled tiles. In this appendix the relevance of this concept to a DEMO is examined with particular emphasis given to the ability of the cooling tube arrangement to remove the heat. A suggested modification to the arrangement of coolant tubes is suggested so that the design can operate at the higher loadings of a DEMO. (author)

Osiris and SOMBRERO inertial confinement fusion power plant designs

International Nuclear Information System (INIS)

Meier, W.R.; Bieri, R.L.; Monsler, M.J.

1992-03-01

The primary objective of the of the IFE Reactor Design Studies was to provide the Office of Fusion Energy with an evaluation of the potential of inertial fusion for electric power production. The term reactor studies is somewhat of a misnomer since these studies included the conceptual design and analysis of all aspects of the IFE power plants: the chambers, heat transport and power conversion systems, other balance of plant facilities, target systems (including the target production, injection, and tracking systems), and the two drivers. The scope of the IFE Reactor Design Studies was quite ambitious. The majority of our effort was spent on the conceptual design of two IFE electric power plants, one using an induction linac heavy ion beam (HIB) driver and the other using a Krypton Fluoride (KrF) laser driver. After the two point designs were developed, they were assessed in terms of their (1) environmental and safety aspects; (2) reliability, availability, and maintainability; (3) technical issues and technology development requirements; and (4) economics. Finally, we compared the design features and the results of the assessments for the two designs

Plasma regimes and research goals of JT-60SA towards ITER and DEMO

International Nuclear Information System (INIS)

Kamada, Y.; Ide, S.; Fujita, T.; Suzuki, T.; Matsunaga, G.; Yoshida, M.; Shinohara, K.; Urano, H.; Nakano, T.; Sakurai, S.; Kawashima, H.; Barabaschi, P.; Lackner, K.; Ishida, S.; Bolzonella, T.

2011-01-01

The JT-60SA device has been designed as a highly shaped large superconducting tokamak with a variety of plasma actuators (heating, current drive, momentum input, stability control coils, resonant magnetic perturbation coils, W-shaped divertor, fuelling, pumping, etc) in order to satisfy the central research needs for ITER and DEMO. In the ITER- and DEMO-relevant plasma parameter regimes and with DEMO-equivalent plasma shapes, JT-60SA quantifies the operation limits, plasma responses and operational margins in terms of MHD stability, plasma transport and confinement, high-energy particle behaviour, pedestal structures, scrape-off layer and divertor characteristics. By integrating advanced studies in these research fields, the project proceeds 'simultaneous and steady-state sustainment of the key performances required for DEMO' with integrated control scenario development applicable to the highly self-regulating burning high-β high bootstrap current fraction plasmas.

Constructing a Tibetan Demos in Exile

DEFF Research Database (Denmark)

Brox, Trine

2012-01-01

homeland. Two specific instances of the construction of a transnational exile demos are investigated: citizenship and political representation. The Tibetan Government-in-Exile's formalized idea of citizenship builds upon ideals of equal and loyal members who form a single unit bounded by a common cause...

Induction of protein body formation in plant leaves by elastin-like polypeptide fusions

Directory of Open Access Journals (Sweden)

Joensuu Jussi J

2009-08-01

Full Text Available Abstract Background Elastin-like polypeptides are synthetic biopolymers composed of a repeating pentapeptide 'VPGXG' sequence that are valuable for the simple non-chromatographic purification of recombinant proteins. In addition, elastin-like polypeptide fusions have been shown to enhance the accumulation of a range of different recombinant proteins in plants, thus addressing the major limitation of plant-based expression systems, which is a low production yield. This study's main objectives were to determine the general utility of elastin-like polypeptide protein fusions in various intracellular compartments and to elucidate elastin-like polypeptide's mechanism of action for increasing recombinant protein accumulation in the endoplasmic reticulum of plants. Results The effect of elastin-like polypeptide fusions on the accumulation of green fluorescent protein targeted to the cytoplasm, chloroplasts, apoplast, and endoplasmic reticulum was evaluated. The endoplasmic reticulum was the only intracellular compartment in which an elastin-like polypeptide tag was shown to significantly enhance recombinant protein accumulation. Interestingly, endoplasmic reticulum-targeted elastin-like polypeptide fusions induced the formation of a novel type of protein body, which may be responsible for elastin-like polypeptide's positive effect on recombinant protein accumulation by excluding the heterologous protein from normal physiological turnover. Although expressed in the leaves of plants, these novel protein bodies appeared similar in size and morphology to the prolamin-based protein bodies naturally found in plant seeds. The elastin-like polypeptide-induced protein bodies were highly mobile organelles, exhibiting various dynamic patterns of movement throughout the cells, which were dependent on intact actin microfilaments and a functional actomyosin motility system. Conclusion An endoplasmic reticulum-targeted elastin-like polypeptide fusion approach

Which are the competitors for a fusion power plant?

International Nuclear Information System (INIS)

Miller, Ronald L.

2000-01-01

The (future) competitive position of central-station fusion power will depend on the resolution of several broad public-policy issues, including the provision of adequate electrical energy to a growing world population and the interaction of economic and environmental considerations meeting evolving standards of public acceptance and regulatory compliance. Candidate baseload central-station power plants, fusion or other, will be expected to contend for preferential market penetration against an evolving set of performance indicators or metrics (e.g. cost of electricity) reflecting societal 'customer preferences' for abundant, affordable, safe, reliable, and environmentally benign sources. This competition is enhanced by transitions to price-deregulated regimes, overlaid by nuclear uncertainites and evolution beyond carbon-based fuels toward more renewables in the energy mix. From these top-level considerations, quantifiable attributes, including plant size (output), system power density, surface heat flux, recirculating power fraction, power-conversion efficiency, waste streams, and forced- and planned-outage rates emerge

Results of availability imposed configuration details developed for K-DEMO

Energy Technology Data Exchange (ETDEWEB)

Brown, Tom, E-mail: tbrown@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Titus, Peter; Brooks, Art; Zhang, Han; Neilson, Hutch [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Im, Kihak; Kim, Keeman [National Fusion Research Institute, Daejeon 305-806 (Korea, Republic of)

2016-11-01

The Korean fusion demonstration reactor (K-DEMO) has completed a two year study looking at key Tokamak components and configuration options in preparation of a conceptual design phase. A key part of a device configuration centers on defining an arrangement that enhances the ability to reach high availability values by defining design solutions that foster simplified maintenance operations. To maximize the size and minimize the number of in-vessel components enlarged TF coils were defined that incorporate a pair of windings within each coil to mitigate pressure drop issues and to reduce the cost of the coils. A semi-permanent shield structure was defined to develop labyrinth interfaces between double-null plasma contoured shield modules, provide an entity to align blanket components and provide support against disruption loads—with a load path that equilibrates blanket, TF and PF loads through a base structure. Blanket piping services and auxiliary systems that interface with in-vessel components have played a major role in defining the overall device arrangement—concept details will be presented along with general arrangement features and preliminary results obtained from disruption analysis.

Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

International Nuclear Information System (INIS)

Chen, Yuming; Ghidersa, Bradut-Eugen; Jin, Xue Zhou

2016-01-01

Highlights: • This paper presents transient CFD analyses on the cooling channels of the DEMO HCPB FW for accidental scenarios LOCA and LOFA. • In both LOCA & LOFA, the wall temperature increases quickly to an unacceptable level within seconds. • If the coolant flow rate is maintained at a half of nominal value in case of LOFA (partial LOFA), the wall temperature rises much slower, but will still leads to a damage of structure within minutes. • The simulated heat transfer coefficients were compared with empirical correlations. - Abstract: Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

Transient analyses on the cooling channels of the DEMO HCPB blanket concept under accidental conditions

Energy Technology Data Exchange (ETDEWEB)

Chen, Yuming, E-mail: Yuming.chen@kit.edu; Ghidersa, Bradut-Eugen; Jin, Xue Zhou

2016-11-01

Highlights: • This paper presents transient CFD analyses on the cooling channels of the DEMO HCPB FW for accidental scenarios LOCA and LOFA. • In both LOCA & LOFA, the wall temperature increases quickly to an unacceptable level within seconds. • If the coolant flow rate is maintained at a half of nominal value in case of LOFA (partial LOFA), the wall temperature rises much slower, but will still leads to a damage of structure within minutes. • The simulated heat transfer coefficients were compared with empirical correlations. - Abstract: Helium Cooled Pebble Bed (HCPB) blanket concept is one of the DEMO (Demonstration Power Plant) blanket concepts running for the final DEMO design selection. In this paper, transient analyses on the cooling channels of the FW are carried out by means of CFD simulations for the selected accidental scenarios loss-of-coolant-accident (LOCA) and loss-of-flow-accident (LOFA). ANSYS-CFX is used for the simulations. The simulation results help to understand how fast the temperature of the FW can increase and what is the time window that is available until the temperature of the structural material reaches the design limit in order to be able to define a suitable protection strategy for the system. In view of later developments of the models, the heat transfer coefficients calculated with CFD are compared with the values predicted by two widely used correlations for turbulent pipe flows.

The nuclear fusion reactor. How close are we to its realisation

International Nuclear Information System (INIS)

Lackner, K.

2001-01-01

A fusion power plant would rely on practically unlimited supplies of primary materials, and possess very favourable environmental and safety properties. Exploiting the nuclear fusion reaction for continuous power production requires, however, the solution of some of the most demanding physics and technology issues. At the same time the final proof of principle of a self-sustaining fusion reaction can only be delivered in a device with a thermal power rating in the 1 GW range, as the power production by fusion reactions increases much stronger with volume than the energy losses from a hot plasma. A range of well conceived tokamak devices have generated during the last couple of decades the experimental basis and the physics understanding for proceeding now to the construction of such a demonstration device: ITER. This device will also incorporate, partly in the form of test modules, nearly all the critical technologies required for the operation of a commercial power plant. A notable exception to this are materials, where the much lower neutron fluence of ITER allows to work with readily available, conventional steels, whereas power plants will have to use radiation resistant, low-activation materials now under development. The presentation summarizes briefly the development path that has led to the ITER design, and the physics criteria determining its layout. Seven technologies were identified by the international design team (constituted by engineers and scientists from Japan, the European Union, the Russian Federation, and the USA) as critical, and made subject to seven large R and D projects, successfully carried out with an investment of about 400 Million $. The roadmap for the development beyond ITER foresees as subsequent step a power plant (DEMO) that will already be largely identical to the first generation of commercial installations. The physics of this device will be completely verified by the beginning of its planning by the operating experience of ITER

Overview of principles and challenges of fusion nuclear technology

International Nuclear Information System (INIS)

Abdou, M.

2007-01-01

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

Concept design of DEMO divertor cassette remote handling: Simply supported beam approach

Energy Technology Data Exchange (ETDEWEB)

Mozzillo, Rocco [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Di Gironimo, Giuseppei, E-mail: peppe.digironimo@gmail.com [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mäkinen, Harri [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Miccichè, Gioacchino [ENEA – CR Brasimone, I-40032 Camugnano, BO (Italy); Määttä, Timo [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2017-03-15

Highlights: • The present work focused on a new approach to the design of DEMO Divertor Cassette Remote Handling Equipment. • The work provides an alternative approach to the design based on the concept of a simply supported beam. • The approach proposed focuses a Divertor Cassette mover that performs the maintenance of the three cassettes at each port. • First rough dimensioning of the main components has been provided and demonstrating the feasibility of the design solutions. • The main idea of the work consisted on a design capable to use knowledge already adopted in industrial contexts. - Abstract: The present work focused on the development of a new approach to the concept design of DEMO Divertor Cassette (DC) Remote Handling Equipment (RHE). The approach is based on three main assumptions: the DC remote handling activities and the equipment shall be simplified as much as possible; technologies well known and consolidated in the industrial context can be adopted also in the nuclear fusion field; the design of the RHE should be based on a simply supported beam approach instead of cantilever approach. In detail, during the maintenance activities the barycentre of the DC is centred with respect to DC supports. This solution could simplify the design of RHE with a consequent reduction of the design and development costs. Moreover also the DC remote handling tasks shall be simplified in order to better manage the DC maintenance processes. For this reason the DC assembly and disassembly process has been simplified dividing the main sequences in basic movements. For each movement a dedicated tool has been conceived.

The Role of Community Colleges in Advancing Upward Mobility: A Demos Perspective

Science.gov (United States)

Huelsman, Mark

2015-01-01

This article provides a short background on Demos, a public policy organization that works on issues of political and economic inequality. Demos views community colleges as a linchpin in the American higher education system, and it has worked over several years to research ways to increase state support for higher education and direct support…

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-03-01

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

Status of development of functional materials with perspective on beyond ITER

International Nuclear Information System (INIS)

Shikama, T.; Knitter, R.; Moeslang, A.; Konys, J.; Deli, L.; Muroga, T.; Kawamura, H.; Kohyama, A.

2007-01-01

Any engineering system is composed of functional materials as well as of structural materials, and more advanced systems tend to demand a more important and versatile role to functional materials. In nuclear fusion systems, examples of principle functional materials will be breeders and neutron multipliers for tritium production, coatings on structural materials for corrosion-resistance, MHD-loss-reduction and control of tritium permeation, thermal insertions for heat transport control, and optical and electrical materials for plasma and environmental diagnostics. For incarnation of a nuclear fusion power plant, namely DEMO, development of the functional materials with appropriate properties is essential. A role of functional materials depends strongly on a specific design of DEMO, namely designs of systems for tritium-breeding, system-cooling and heat-transfer. In the framework of ITER project, development of tritium blanket modules (TBM) is underway. Also, in parallel with the ITER project, a complemental program called the BA (Broader Approach) is launched for realization of a DEMO nuclear fusion reactor in an appropriate time schedule, where key issues of the nuclear fusion engineering needed for the DEMO will be studied under EU/Japan collaboration. In the meantime, technologies and materials needed for diagnostics and control of burning plasma are extensively discussed under the framework of International Tokamak Physics Activity (ITPA). The present paper reviews a present status of development of functional materials from views of internationally coordinated activities based on fundamental aspects of the DEMO demands as well as from views of activities based on specific but currently dominant DEMO designs. Examples of functional materials reviewed here are solid breeders, beryllium and beryllium alloys, coating layers on structural materials, thermal inserts, and some electrical and optical materials. (orig.)

Possible futures for the development of a fusion demonstration plant

International Nuclear Information System (INIS)

Nichols, S.P.

1976-01-01

As indicated by the Fusion Planning Bulletins, the Division of Controlled Thermonuclear Research is becoming involved in planning with alternative scenarios. The Center for Energy Studies at the University of Texas at Austin has been involved with such planning for several years and has examined various scenarios for fusion power development using the Partitive Analytical Forecasting (PAF) technique. The most recent studies compare the long-term plan presented in WASH-1290, Fusion Power by Magnetic Confinement, with other plans that have been proposed, such as the plan proposed by Kulcinski and Conn of the University of Wisconsin. The study indicates that some of the alternative plans do have possibilities to shorten the required time for the completion of a demonstration fusion plant without increased costs or a decrease in the likelihood of success. The current efforts of the project are in the planning exercises recently completed by committees set up by the DCTR. Further comparisons of alternative scenarios will be performed as part of this effort

Status of advanced tritium breeder development for DEMO in the broader approach activities in Japan

International Nuclear Information System (INIS)

Hoshino, Tsuyoshi; Oikawa, Fumiaki; Nishitani, Takeo

2010-01-01

DEMO reactors require ' 6 Li-enriched ceramic tritium breeders' which have high tritium breeding ratios (TBRs) in the blanket designs of both EU and JA. Both parties have been promoting the development of fabrication technologies of Li 2 TiO 3 pebbles and of Li 4 SiO 4 pebbles including the reprocessing. However, the fabrication techniques of tritium breeders pebbles have not been established for large quantities. Therefore, these parties launch a collaborative project on scaleable and reliable production routes of advanced tritium breeders. In addition, this project aims to develop fabrication techniques allowing effective reprocessing of 6 Li. The development of the production and 6 Li reprocessing techniques includes preliminary fabrication tests of breeder pebbles, reprocessing of lithium, and suitable out-of-pile characterizations. The R and D on the fabrication technologies of the advanced tritium breeders and the characterization of developed materials has been started between the EU and Japan in the DEMO R and D of the International Fusion Energy Research Centre (IFERC) project as a part of the Broader Approach activities from 2007 to 2016. The equipment for production of advanced breeder pebbles is planned will be installed in the DEMO R and D building at Rokkasho, Japan. The design work in this facility was carried out. The specifications of the pebble production apparatuses and related equipment in this facility were fixed, and the basic data of these apparatuses was obtained. In this design work, the preliminary investigations of the dissolution and purification process of tritium breeders were carried out. From the results of the preliminary investigations, lithium resources of 90% above were recovered by the aqueous dissolving methods using HNO 3 and H 2 O 2 . The removal efficiency of 60 Co by the addition in the dissolved solutions of lithium ceramics were 97-99.9% above using activated carbon impregnated with 8-hydroxyquinolinol. In this report

A system dynamics model for stock and flow of tritium in fusion power plant

Energy Technology Data Exchange (ETDEWEB)

Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kwon, Saerom [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Sakamoto, Yoshiteru; Yamanishi, Toshihiko; Tobita, Kenji [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori-ken 039-3212 (Japan)

2015-10-15

Highlights: • System dynamics model of tritium fuel cycle was developed for analyzing stock and flow of tritium in fusion power plants. • Sensitivity of tritium build-up to breeding ratio parameters has been assessed to two plant concepts having 3 GW and 1.5 GW fusion power. • D-D start-up absolutely without initial loading of tritium is possible for both of the 3 GW and 1.5 GW fusion power plant concepts. • Excess stock of tritium is generated by the steady state operation with the value of tritium breeding ratio over unity. - Abstract: In order to analyze self-efficiency of tritium fuel cycle (TFC) and share the systems thinking of TFC among researchers and engineers in the vast area of fusion reactor technology, we develop a system dynamics (SD) TFC model using a commercial software STELLA. The SD-TFC model is illustrated as a pipe diagram which consists of tritium stocks, such as plasma, fuel clean up, isotope separation, fueling with storage and blanket, and pipes connecting among them. By using this model, we survey a possibility of D-D start-up without initial loading of tritium on two kinds of fusion plant having different plasma parameters. The D-D start-up scenario can reduce the necessity of initial loading of tritium through the production in plasma by D-D reaction and in breeding blanket by D-D neutron. The model is also used for considering operation scenario to avoid excess stock of tritium which must be produced at tritium breeding ratio over unity.

The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

International Nuclear Information System (INIS)

Fischer, U.; Avrigeanu, M.; Avrigeanu, V.; Cabellos, O.; Kodeli, I.; Koning, A.; Konobeyev, A.Yu.; Leeb, H.; Rochman, D.; Pereslavtsev, P.; Sauvan, P.; Sublet, J.-C.; Trkov, A.; Dupont, E.; Leichtle, D.; Izquierdo, J.

2014-01-01

This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations

The Activities of the European Consortium on Nuclear Data Development and Analysis for Fusion

Energy Technology Data Exchange (ETDEWEB)

Fischer, U., E-mail: ulrich.fischer@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Avrigeanu, M.; Avrigeanu, V. [Horia Hulubei National Institute of Physics and Nuclear Engineering (IFIN-HH), RO-077125 Magurele (Romania); Cabellos, O. [Departamento de Ingenieria Nuclear, Universidad Politecnica de Madrid, 28006 Madrid (Spain); Kodeli, I. [Jozef Stefan Institute (JSI), Jamova 39, 1000 Ljubljana (Slovenia); Koning, A. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 LE Petten (Netherlands); Konobeyev, A.Yu. [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Leeb, H. [Technische Universitaet Wien, Atominstitut, Wiedner Hauptstrasse 8–10, 1040 Wien (Austria); Rochman, D. [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 LE Petten (Netherlands); Pereslavtsev, P. [Karlsruhe Institute of Technology, Institute for Neutron Physic and Reactor Technology, 76344 Eggenstein-Leopoldshafen (Germany); Sauvan, P. [Universidad Nacional de Educacion a Distancia, C. Juan del Rosal, 12, 28040 Madrid (Spain); Sublet, J.-C. [Euratom/CCFE Fusion Association, Culham Science Centre, OX14 3DB (United Kingdom); Trkov, A. [Jozef Stefan Institute (JSI), Jamova 39, 1000 Ljubljana (Slovenia); Dupont, E. [OECD Nuclear Energy Agency, Paris (France); Leichtle, D.; Izquierdo, J. [Fusion for Energy, Barcelona (Spain)

2014-06-15

This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.

Japanese contributions to ITER testing program of solid breeder blankets for DEMO

International Nuclear Information System (INIS)

Kuroda, Toshimasa; Yoshida, Hiroshi; Takatsu, Hideyuki; Maki, Koichi; Mori, Seiji; Kobayashi, Takeshi; Suzuki, Tatsushi; Hirata, Shingo; Miura, Hidenori.

1991-04-01

ITER Conceptual Design Activity (CDA), which has been conducted by four parties (Japan, EC, USA and USSR) since May 1988, has been finished on December 1990 with a great achievement of international design work of the integrated fusion experimental reactor. Numerous issues of physics and technology have been clarified for providing a framework of the next phase of ITER (Engineering Design Activity; EDA). Establishment of an ITER testing program, which includes technical test issues of neutronics, solid breeder blankets, liquid breeder blankets, plasma facing components, and materials, has been one of the goals of the CDA. This report describes Japanese proposal for the testing program of DEMO/power reactor blanket development. For two concepts of solid breeder blanket (helium-cooled and water-cooled), identification of technical issues, scheduling of test program, and conceptual design of test modules including required test facility such as cooling and tritium recovery systems have been carried out as the Japanese contribution to the CDA. (author)

Overview of the present progress and activities on the CFETR

Science.gov (United States)

Wan, Yuanxi; Li, Jiangang; Liu, Yong; Wang, Xiaolin; Chan, Vincent; Chen, Changan; Duan, Xuru; Fu, Peng; Gao, Xiang; Feng, Kaiming; Liu, Songlin; Song, Yuntao; Weng, Peide; Wan, Baonian; Wan, Farong; Wang, Heyi; Wu, Songtao; Ye, Minyou; Yang, Qingwei; Zheng, Guoyao; Zhuang, Ge; Li, Qiang; CFETR Team

2017-10-01

The China Fusion Engineering Test Reactor (CFETR) is the next device in the roadmap for the realization of fusion energy in China, which aims to bridge the gaps between the fusion experimental reactor ITER and the demonstration reactor (DEMO). CFETR will be operated in two phases. Steady-state operation and self-sufficiency will be the two key issues for Phase I with a modest fusion power of up to 200 MW. Phase II aims for DEMO validation with a fusion power over 1 GW. Advanced H-mode physics, high magnetic fields up to 7 T, high frequency electron cyclotron resonance heating and lower hybrid current drive together with off-axis negative-ion neutral beam injection will be developed for achieving steady-state advanced operation. The recent detailed design, research and development (R&D) activities including integrated modeling of operation scenarios, high field magnet, material, tritium plant, remote handling and future plans are introduced in this paper.

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.

Annual report of Fusion Research and Development Directorate of JAEA for FY2008 and FY2009

International Nuclear Information System (INIS)

Isei, Nobuaki

2011-03-01

This annual report provides an overview of major results and progress on research and development (R and D) activities at Fusion Research and Development Directorate of Japan Atomic Energy Agency (JAEA) for FY2008 (from April 1, 2008 to March 31, 2009) and FY2009 (from April 1, 2009 to March 31, 2010), including those performed in collaboration with other research establishments of JAEA, research institutes, and universities. Concerning the ITER project, JAEA was nominated as the domestic agency by the Japanese government after the ITER Agreement took effect, and has fulfilled the obligations. In the development of superconducting conductors, JAEA constructed a technical platform for the fabrication of superconducting conductors for toroidal field (TF) coils ahead of other countries. JAEA immediately started and completed the construction of a plant to fabricate superconducting conductors, and started their fabrication ahead of other countries. In the development of gyrotron high-frequency heating equipment, since only the JAEA satisfies the ITER's procurement specifications among supplier countries, the ITER Organization requested JAEA to conduct confidence tests, and achieved results such as data acquisition that could contribute to the development of the ITER's operational scenario. For the development of neutral beam injectors, advantages of the multi-stage acceleration system developed by JAEA was recognized as a result of comparative experiments with single-stage acceleration systems developed in Europe for the particle acceleration system, and was adopted in the ITER's technical specifications. For the Broader Approach (BA) activities, JAEA was designated as the implementing agency by the Japanese government after the BA Agreement took effect, and has fulfilled the obligations and promoted three projects in the BA activities steadily through domestic cooperation and coordination with Europe. Concerning activities related to the International Fusion Energy

Pre-conceptual design assessment of DEMO remote maintenance

Energy Technology Data Exchange (ETDEWEB)

Loving, A., E-mail: antony.loving@ccfe.ac.uk [EURATOM/Culham Center Fusion Energy, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Crofts, O.; Sykes, N.; Iglesias, D.; Coleman, M.; Thomas, J. [EURATOM/Culham Center Fusion Energy, Culham Science Centre, OX14 3DB, Abingdon (United Kingdom); Harman, J. [EFDA Close Support Unit Garching, Boltzmannstaße 2, D-85748 Garching Bei München (Germany); Fischer, U. [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Sanz, J. [Instituto de Fusión Nuclear/UPM, Madrid (Spain); Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Mittwollen, M. [Karlsruhe Institute of Technology, Institut für Fördertechnik und Logistiksysteme, Gotthard-Franz-Straße 8, Geb.50.38, 76131 Karlsruhe (Germany)

2014-10-15

EDFA, as part of the Power Plant Physics and Technology programme, has been working on the pre-conceptual design of a Demonstration Power Plant (DEMO). As part of this programme, a review of the remote maintenance strategy considered maintenance solutions compatible with expected environmental conditions, whilst showing potential for meeting the plant availability targets. A key finding was that, for practical purposes, the expected radiation levels prohibit the use of complex remote handling operations to replace the first wall. In 2012/2013, these remote maintenance activities were further extended, providing an insight into the requirements, constraints and challenges. In particular, the assessment of blanket and divertor maintenance, in light of the expected radiation conditions and availability, has elaborated the need for a very different approach from that of ITER. This activity has produced some very informative virtual reality simulations of the blanket segments and pipe removal that are exceptionally valuable in communicating the complexity and scale of the required operations. Through these simulations, estimates of the maintenance task durations have been possible demonstrating that a full replacement of the blankets within 6 months could be achieved. The design of the first wall, including the need to use sacrificial limiters must still be investigated. In support of the maintenance operations, a first indication of the requirements of an Active Maintenance Facility (AMF) has been elaborated.

The heat transport system and plant design for the HYLIFE-2 fusion reactor

International Nuclear Information System (INIS)

Hoffman, M.A.

1990-01-01

HYLIFE is the name given to a family of self-healing liquid-wall reactor concepts for inertial confinement fusion. This HYLIFE-II concept employs the molten salt, Flibe, for the liquid jets instead of liquid lithium used in the original HYLIFE-I study. A preliminary conceptual design study of the heat transport system and the balance of plant of the HYLIFE-II fusion power plant is described in this paper with special emphasis on a scoping study to determine the best intermediate heat exchanger geometry and flow conditions for minimum cost of electricity. 11 refs., 8 figs

Introduction: Undoing the demos

DEFF Research Database (Denmark)

Dean, Mitchell

2017-01-01

and usefulness of Michel Foucault’s notion of governmentality and Karl Marx’s analysis of capitalism for analysing neoliberalism; the way that neoliberalism ‘economises’ everything including politics and democracy; the nature of the state and of sovereignty, and how the left should relate to these......; and the nature of critique in its different forms (Kantian, Foucauldian, Marxist and others). These are issues that are important not only for the specific argument of Undoing the Demos, but more generally for social and political theory today....

Tritium instrumentation for a fusion reactor power plant

International Nuclear Information System (INIS)

Shank, K.E.; Easterly, C.E.

1976-09-01

A review of tritium instrumentation is presented. This includes a discussion of currently available in-plant instrumentation and methods required for sampling stacks, monitoring process streams and reactor coolants, analyzing occupational work areas for air and surface contamination, and personnel monitoring. Off-site instrumentation and collection techniques are also presented. Conclusions are made concerning the adequacy of existing instrumentation in relation to the monitoring needs for fusion reactors

The DEMO wall load challenge

Czech Academy of Sciences Publication Activity Database

Wenninger, R.; Albanese, R.; Ambrosino, R.; Arbeiter, F.; Aubert, J.; Bachmann, C.; Barbato, L.; Barrett, T.; Beckers, M.; Biel, W.; Boccaccini, L.; Carralero, D.; Coster, D.; Eich, T.; Fasoli, A.; Federici, G.; Firdaouss, M.; Graves, J.; Horáček, Jan; Kovari, M.; Lanthaler, S.; Loschiavo, V.; Lowry, C.; Lux, H.; Maddaluno, G.; Maviglia, F.; Mitteau, R.; Neu, R.; Pfefferle, D.; Schmid, K.; Siccinio, M.; Sieglin, B.; Silva, C.; Snicker, A.; Subba, F.; Varje, J.; Zohm, H.

2017-01-01

Roč. 57, č. 4 (2017), č. článku 046002. ISSN 0029-5515 EU Projects: European Commission(XE) 633053 - EUROfusion Institutional support: RVO:61389021 Keywords : DEMO * power loads * first wall Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa4fb4

Demos as an Explanatory Lens in Teacher Educators' Elusive Search for Social Justice

Science.gov (United States)

Oikonomidoy, Eleni M.; Brock, Cynthia H.; Obenchain, Kathryn M.; Pennington, Julie L.

2013-01-01

Borrowing insights from the Ancient Greek ideal conceptions of a democratic civic space (demos), this article examines the applicability of this framework to four teacher educators' journey to implement social justice in their programs. It is proposed that the three constitutive dimensions of demos (freedom of speech, equality to vote and hold…

Laser requirements for a laser fusion energy power plant

Institute of Scientific and Technical Information of China (English)

Stephen; E.Bodner; Andrew; J.Schmitt; John; D.Sethian

2013-01-01

We will review some of the requirements for a laser that would be used with a laser fusion energy power plant, including frequency, spatial beam smoothing, bandwidth, temporal pulse shaping, efficiency, repetition rate, and reliability. The lowest risk and optimum approach uses a krypton fluoride gas laser. A diode-pumped solid-state laser is a possible contender.

European development of He-cooled divertors for fusion power plants

International Nuclear Information System (INIS)

Norajitra, P.; Giniyatulin, R.; Kuznetsov, V.; Mazul, I.; Ovchinnikov, I.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Karditsas, P.; Maisonnier, D.; Sardain, P.; Nardi, C.; Papastergiou, S.; Pizzuto, A.

2005-01-01

Helium-cooled divertor concepts are considered suitable for use in fusion power plants for safety reasons, as they enable the use of a coolant compatible with any blanket concept, since water would not be acceptable e.g. in connection with ceramic breeder blankets using large amounts of beryllium. Moreover, they allow for a high coolant exit temperature for increasing the efficiency of the power conversion system. Within the framework of the European power plant conceptual study (PPCS), different helium-cooled divertor concepts based on different heat transfer mechanisms are being investigated at ENEA Frascati, Italy, and Forschungszentrum Karlsruhe, Germany. They are based on a modular design which helps reduce thermal stresses. The design goal is to withstand a high heat flux of about 10-15 MW/m 2 , a value which is considered relevant to future fusion power plants to be built after ITER. The development and optimisation of the divertor concepts require an iterative design approach with analyses, studies of materials and fabrication technologies, and the execution of experiments. These issues and the state of the art of divertor development shall be the subject of this report. (author)

Radiation effects on superconducting fusion magnet components

International Nuclear Information System (INIS)

Weber, H.W.

2011-01-01

Nuclear fusion devices based on the magnetic confinement principle heavily rely on the existence and performance of superconducting magnets and have always significantly contributed to advancing superconductor and magnet technology to their limits. In view of the presently ongoing construction of the tokamak device ITER and the stellerator device Wendelstein 7X and their record breaking parameters concerning size, complexity of design, stored energy, amperage, mechanical and magnetic forces, critical current densities and stability requirements, it is deemed timely to review another critical parameter that is practically unique to these devices, namely the radiation response of all magnet components to the lifetime fluence of fast neutrons and gamma rays produced by the fusion reactions of deuterium and tritium. I will review these radiation effects in turn for the currently employed standard "technical" low temperature superconductors NbTi and Nb 3 Sn, the stabilizing material (Cu) as well as the magnet insulation materials and conclude by discussing the potential of high temperature superconducting materials for future generations of fusion devices, such as DEMO. (author)

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

Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

Science.gov (United States)

Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

2017-12-01

Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part

“PROCESS”: A systems code for fusion power plants—Part 1: Physics

Energy Technology Data Exchange (ETDEWEB)

Kovari, M., E-mail: michael.kovari@ccfe.ac.uk; Kemp, R.; Lux, H.; Knight, P.; Morris, J.; Ward, D.J.

2014-12-15

Highlights: • PROCESS is a fusion reactor systems code. • It optimises a figure of merit subject to constraints chosen by the user. • CCFE are working to make the assumptions and equations explicit and public. • The PROCESS homepage is (www.ccfe.ac.uk/powerplants.aspx). - Abstract: PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system, from the basic plasma physics to the generation of electricity. It has been used for many years, but details of its operation have not been previously published. This paper describes some of its capabilities. PROCESS is usually used in optimisation mode, in which it finds a set of parameters that maximise (or minimise) a figure of merit chosen by the user, while being consistent with the inputs and the specified constraints. Because the user can apply all the physically relevant constraints, while allowing a large number of parameters to vary, it is in principle only necessary to run the code once to produce a self-consistent, physically plausible reactor model. The scope of PROCESS is very wide and goes well beyond reactor physics, including conversion of heat to electricity, buildings, and costs, but this paper describes only the plasma physics and magnetic field calculations. The capabilities of PROCESS in plasma physics are limited, as its main aim is to combine engineering, physics and economics. A model is described which shows the main plasma features of an inductive ITER scenario. Significant differences between the PROCESS results and the published scenario include the bootstrap current and loop voltage. The PROCESS models for these are being revised. Two new models for DEMO have been obtained. The first, DEMO A, is intended to be “conservative” in that it might be possible to build it using the technology of the near future. For example, since current drive technologies are not yet

On tungsten technologies and qualification for DEMO

International Nuclear Information System (INIS)

Laan, J. van der; Hegeman, H.; Wouters, O.; Luzginova, N.; Jonker, B.; Van der Marck, S.; Opschoor, J.; Wang, J.; Dowling, G.; Stuivenga, M.; Carton, E.

2009-01-01

Tungsten alloys are considered prime candidates for the in-vessel components directly facing the plasma. For example, in the HEMJ helium cooled divertor design tiles may be operated at temperatures up to 1700 deg. C, supported by a structure partially consisting of tungsten at temperatures from 600 to 1000 deg. C, and connected to a HT steel structure. The tungsten armoured primary wall is operated at 500-900 deg. C. Irradiation doses will be few tens dpa at minimum, but FPR requirements for plants availability will stretch these targets. Recently injection moulding technology was developed for pure tungsten and representative parts were manufactured for ITER monobloc divertors and DEMO HEMJ thimbles. The major advantages for this technology are the efficient use of material feedstock/resources and the intrinsic possibility to produce near-finished product, avoiding machining processes that are costly and may introduce surface defects deteriorating the component in service performance. It is well suited for mass-manufacturing of components as well known in e.g. lighting industries. To further qualify this material technology various specimen types were produced with processing parameters identical to the components, and tested successfully, showing the high potential for implementation in (fusion) devices. Furthermore, the engineering approach can clearly be tailored away from conventional design and manufacturing technologies based on bulk materials. The technology is suitable for shaping of new W-alloys and W-ODS variants as well. Basically this technology allows a particular qualification trajectory. There is no need to produce large batches of material during the material development and optimization stage. For the verification of irradiation behaviour in the specific neutron spectra, there is a further attractive feature to use e.g. isotope tailored powders to adjust to available irradiation facilities like MTR's. In addition the ingrowth of transmutation

Divertor development for a future fusion power plant

International Nuclear Information System (INIS)

Norajitra, Prachai

2011-01-01

Nuclear fusion is considered as a future source of sustainable energy supply. In the first chapter, the physical principle of magnetic plasma confinement, and the function of a tokamak are described. Since the discovery of the H-mode in ASDEX experiment ''Divertor I'' in 1982, the divertor has been an integral part of all modern tokamaks and stellarators, not least the ITER machine. The goal of this work is to develop a feasible divertor design for a fusion power plant to be built after ITER. This task is particularly challenging because a fusion power plant formulates much greater demands on the structural material and the design than ITER in terms of neutron wall load and radiation. First several divertor concepts proposed in the literature e.g. the Power Plant Conceptual Study (PPCS) using different coolants are reviewed and analyzed with respect to their performance. As a result helium cooled divertor concept exhibited the best potential to come up to the highest safety requirements and therefore has been chosen for the design process. From the third chapter the necessary steps towards this goal are described. First, the boundary conditions for the arrangement of a divertor with respect to the fusion plasma are discussed, as this determines the main thermal and neutronic load parameters. Based on the loads material selection criteria are inherently formulated. In the next step, the reference design is defined in accordance with the established functional design specifications. The developed concept is of modular nature and consists of cooling fingers of tungsten using an impingement cooling in order to achieve a heat dissipation of 10 MW/m 2 . In the next step, the design was subjected to the thermal-hydraulic and thermo-mechanical calculations in order to analyze and improve the performance and the manufacturing technologies. Based on these results, a prototype was produced and experimentally tested on their cooling capacity, their thermo-cyclic loading

Systems studies of dual purpose electric/synthetic fuels fusion plants

International Nuclear Information System (INIS)

Beardsworth, E.; Powell, J.

1975-02-01

A reactor power plant is proposed that can meet base load electrical demand, while the remainder can generate synthetic fuels and meet intermittent electrical demands. Two principal objectives of this study are: (1) to examine how strongly various economic demand and resource factors affect the amount of installed CTR capacity, and (2) to examine what increase in CTR capacity can be expected with dual purpose electric/synthetic fuel fusion plants, and also the relative importance of the different production modes

Tritiated hydrogen gas storage systems for a fusion plant

International Nuclear Information System (INIS)

Bramy, W.; Hircq, B.; Peyrat, M.; Leger, D.

1992-01-01

This paper reports that USSI INGENIERIE has carried out a study financed by European Communities Commission concerning the NET/ITER project, on tritium Fuel Management and Storage systems of the International Thermonuclear Experimental Reactor. A processing block diagram for hydrogen isotopes represents all interfaces and possible links between these systems and tritiated gas mixtures flowing through the Fusion plant. Large quantities of hydrogen isotopes (up to several thousand moles of protium, deuterium and tritium) in gaseous form associated with torus fuelling and exhaust pellet injection, and neutral beam injection, must be stored and managed in such a plant

Progress of IFERC project in the Broader Approach Activities

International Nuclear Information System (INIS)

Araki, M.; Sakamoto, Y.; Hayashi, K.; Ohuchi, R.; Nishitani, T.

2010-01-01

For contributing to the ITER project and promoting a possible early realization of DEMO, since 2007 the International Fusion Energy Research Centre (IFERC) shall perform the activities on (1) DEMO Design and R and D Coordination, (2) Computational Simulation Centre, and (3) ITER Remote Experimentation Centre in the framework under the agreement between the Government of Japan and the European Atomic Energy Community for the joint implementation of the Broader Approach Activities in the field of fusion energy. The DEMO Design activity aims at establishing a common basis for DEMO Design, including design features of DEMO, possible common concepts of DEMO, a roadmap for DEMO, and so on. Based on the common interest of EU and Japan toward DEMO, the DEMO R and D activities have been planned and carried out for the areas which are related to DEMO blanket development: SiC/SiC composites, tritium technology, materials engineering, advanced neutron multiplier, and advanced tritium breeders for DEMO blanket. In the activity of the Computational Simulation Centre, the objective is to provide and exploit a supercomputer for large scale simulation activities to analyse experimental data on fusion plasmas, prepare scenarios for ITER operation, predict the performance of the ITER facilities and contribute to the DEMO Design.