WorldWideScience

Sample records for fusion neutron wall

  1. Characterization for fusion first-wall damage studies of using tailored D-T neutron fields

    International Nuclear Information System (INIS)

    Dierckx, R.; Emigh, C.R.

    1979-01-01

    The approximation required to apply the Bullough-Haynes results to the present calculations is somewhat crude and may imply that the details of the results contain considerable error. However, when the results for each neutron source are viewed in a relative context, several valid and important observations can be made. The almost identical swelling results obtained for the intense neutron source (INS) with a standard blanket and the fusion first wall are most striking. A further comparison with a fusion reactor shows that even the spatial and energy distributions of the neutron flux are similar. In both the INS with blanket and at the first wall of a fusion reactor, there is a radial source flux component of 14-MeV neutrons and a more or less isotropic flux component of low energy (< 14-MeV) neutrons. One must therefore conclude that from the point-of-view of neutron radiation damage, the INS with a blanket, unlike all other types of neutron sources, is not a simulation environment. It is, in fact, a small scale fusion device, and data obtained from INS irradiation experiments would represent fusion reactor results. Such data could then be used to develop correlative procedures for applying data obtained from other simulation sources to fusion reactor conditions

  2. First wall material damage induced by fusion-fission neutron environment

    Energy Technology Data Exchange (ETDEWEB)

    Khripunov, Vladimir, E-mail: Khripunov_VI@nrcki.ru

    2016-11-01

    Highlights: • The highest damage and gas production rates are experienced within the first wall materials of a hybrid fusion-fission system. • About ∼2 times higher dpa and 4–5 higher He appm are expected compared to the values distinctive for a pure fusion system at the same DT-neutron wall loading. • The specific nuclear heating may be increased by a factor of ∼8–9 due to fusion and fission neutrons radiation capture in metal components of the first wall. - Abstract: Neutronic performance and inventory analyses were conducted to quantify the damage and gas production rates in candidate materials when used in a fusion-fission hybrid system first wall (FW). The structural materials considered are austenitic SS, Cu-alloy and V- alloys. Plasma facing materials included Be, and CFC composite and W. It is shown that the highest damage rates and gas particles production in materials are experienced within the FW region of a hybrid similar to a pure fusion system. They are greatly influenced by a combined neutron energy spectrum formed by the two-component fusion-fission neutron source in front of the FW and in a subcritical fission blanket behind. These characteristics are non-linear functions of the fission neutron source intensity. Atomic displacement damage production rate in the FW materials of a subcritical system (at the safe subcriticality limit of ∼0.95 and the neutron multiplication factor of ∼20) is almost ∼2 times higher compared to the values distinctive for a pure fusion system at the same 14 MeV neutron FW loading. Both hydrogen (H) and helium (He) gas production rates are practically on the same level except of about ∼4–5 times higher He-production in austenitic and reduced activation ferritic martensitic steels. A proper simulation of the damage environment in hybrid systems is required to evaluate the expected material performance and the structural component residence times.

  3. Fusion neutronics

    CERN Document Server

    Wu, Yican

    2017-01-01

    This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

  4. 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)

  5. Intense fusion neutron sources

    International Nuclear Information System (INIS)

    Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.

    2010-01-01

    The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 10 15 -10 21 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 10 20 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

  6. Intense fusion neutron sources

    Science.gov (United States)

    Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.

    2010-04-01

    The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 1015-1021 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 1020 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

  7. Fusion: first wall problems

    International Nuclear Information System (INIS)

    Behrisch, R.

    1976-01-01

    Some of the relevant elementary atomic processes which are expected to be of significance to the first wall of a fusion reactor are reviewed. Up to the present, most investigations have been performed at relatively high ion energies, typically E greater than 5 keV, and even in this range the available data are very poor. If the plasma wall interaction takes place at energies of E greater than 1 keV the impurity introduction and first wall erosion which will take place predominantly by sputtering, will be large and may severely limit the burning time of the plasma. The wall bombardment and surface erosion will presumably not decrease substantially by introducing a divertor. The erosion can only be kept low if the energy of the bombarding ions and neutrals can be kept below the threshold for sputtering of 1 to 10 eV. 93 refs

  8. Neutrons and fusion

    International Nuclear Information System (INIS)

    Maynard, C.W.

    1976-01-01

    The production of energy from fusion reactions does not require neutrons in the fundamental sense that they are required in a fission reactor. Nevertheless, the dominant fusion reaction, that between deuterium and tritium, yields a 14 MeV neutron. To contrast a fusion reactor based on this reaction with the fission case, 3 x 10 20 such neutrons produced per gigawatt of power. This is four times as many neutrons as in an equivalent fission reactor and they carry seven times the energy of the fission neutrons. Thus, they dominate the energy recovery problem and create technological problems comparable to the original plasma confinement problem as far as a practical power producing device is concerned. Further contrasts of the fusion and fission cases are presented to establish the general role of neutrons in fusion devices. Details of the energy deposition processes are discussed and those reactions necessary for producing additional tritium are outlined. The relatively high energy flux with its large intensity will activate almost any materials of which the reactor may be composed. This activation is examined from the point of view of decay heat, radiological safety, and long-term storage. In addition, a discussion of the deleterious effects of neutron interactions on materials is given in some detail; this includes the helium and hydrogen producing reactions and displacement rate of the lattice atoms. The various materials that have been proposed for structural purposes, for breeding, reflecting, and moderating neutrons, and for radiation shielding are reviewed from the nuclear standpoint. The specific reactions of interest are taken up for various materials and finally a report is given on the status and prospects of data for fusion studies

  9. Stellarator fusion neutronics research in Australia

    International Nuclear Information System (INIS)

    Zimin, S.; Cross, R.C.

    1997-01-01

    The new status of the H-INF Heliac Stellaralor as a National Facility and the signed international Implementing Agreement on 'Collaboration in the Development of the Stellarator Concept' represents a significant encouragement for further fusion research in Australia. In this report the future of fusion research in Australia is discussed with special attention being paid to the importance of Stellarator power plant studies and in particular stellarator fusion neutronics. The main differences between tokamak and stellarator neutronics analyses are identified, namely the neutron wall loading, geometrical modelling and total heating in in-vessel reactor components including toroidal field (TF) coils. Due to the more complicated nature of stellarator neutronics analyses, simplified approaches to fusion neutronics already developed for tokamaks are expected to be even more important and widely used for designing a Conceptual Stellarator Power Plant

  10. Summary report of the IAEA advisory group meeting on nuclear data for neutron multiplication in fusion-reactor first-wall and blanket materials

    International Nuclear Information System (INIS)

    Muir, D.W.; Pashchenko, A.B.

    1992-09-01

    The present Report contains the Summary of the IAEA Advisory Group Meeting on Nuclear Data for Neutron Multiplication in Fusion-Reactor First-Wall and Blanket Materials, which was hosted by the Southwest Institute of Nuclear physics and Chemistry (SWINPC) at Chengdu, China and held from 19-21 November 1990. This AGM was organized by the IAEA Nuclear Data Section (NDS), with the cooperation and assistance of local organizers at the SWINPC. The papers which the participants prepared for and presented at the meeting will be published as an INDC report. (author)

  11. Nuclear fusion and neutron processes

    International Nuclear Information System (INIS)

    Orlov, V.V.; Shatalov, G.E.; Sherstnev, K.E.

    1984-01-01

    Problems of providing development of the design of an experimental fusion reactor with necessary neutron-physical data are discussed. Isotope composition of spent fuel in the blanket of a hybride fusion reactor (HFR) is given. Neutron balance of the reactor with Li-blanket and neutron balance of the reactor with Pb-multiplier are disclosed. A simplified scheme of neutron and energy balance in the HFR blanket is given. Development and construction of the experimental power reactor is shown to become the nearest problem of the UTS program. Alongside with other complex physical and technical problems solution of this problem requires realization of a wide program of neutron-physical investigations including measurements with required accuracy of neutron cross sections, development of methodical, program and constant basis of neutron calculations and macroscopic experiments on neutron sources

  12. Overview of Australian activities of fusion neutronics

    International Nuclear Information System (INIS)

    Zimin, S.; Dewar, R.L.

    1999-01-01

    The new status of the H-1NF heliac stellarator as a national facility and the signed international implementing agreement on collaboration in the development of the stellarator concept should together be a significant encouragement for further fusion research in Australia. In this report the future of fusion research in Australia is discussed with special attention being paid to the importance of stellarator power plant studies and in particular stellarator fusion neutronics. The main differences between tokamak and stellarator neutronics analyses are identified, namely the neutron wall loading, geometrical modelling and total heating in in-vessel reactor components including toroidal field (TF) coils. An approach to stellarator (TF) coils heating calculations is discussed. This approach is a modification of a previously reported method of total heating calculations in tokamak TF coils. Due to the more complicated nature of stellarator neutronics analyses, simplified approaches to fusion neutronics already developed for tokamaks are expected to be even more important and widely used for designing a conceptual stellarator power plant. (orig.)

  13. 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.)

  14. Local wall power loading variations in thermonuclear fusion devices

    International Nuclear Information System (INIS)

    Carroll, M.C.; Miley, G.H.

    1989-01-01

    A 2 1/2-dimensional geometric model is presented that allows calculation of power loadings at various points on the first wall of a thermonuclear fusion device. Given average wall power loadings for brems-strahlung, cyclotron radiation charged particles, and neutrons, which are determined from various plasma-physics computation models, local wall heat loads are calculated by partitioning the plasma volume and surface into cells and superimposing the heating effects of the individual cells on selected first-wall differential areas. Heat loads from the entire plasma are thus determined as a function of position on the first-wall surface. Significant differences in local power loadings were found for most fusion designs, and it was therefore concluded that the effect of local power loading variations must be taken into account when calculating temperatures and heat transfer rates in fusion device first walls

  15. Neutron cross sections for fusion

    International Nuclear Information System (INIS)

    Haight, R.C.

    1979-10-01

    First generation fusion reactors will most likely be based on the 3 H(d,n) 4 He reaction, which produces 14-MeV neutrons. In these reactors, both the number of neutrons and the average neutron energy will be significantly higher than for fission reactors of the same power. Accurate neutron cross section data are therefore of great importance. They are needed in present conceptual designs to calculate neutron transport, energy deposition, nuclear transmutation including tritium breeding and activation, and radiation damage. They are also needed for the interpretation of radiation damage experiments, some of which use neutrons up to 40 MeV. In addition, certain diagnostic measurements of plasma experiments require nuclear cross sections. The quality of currently available data for these applications will be reviewed and current experimental programs will be outlined. The utility of nuclear models to provide these data also will be discussed. 65 references

  16. Fusion neutronics experiments and analysis

    International Nuclear Information System (INIS)

    1992-01-01

    UCLA has led the neutronics R ampersand D effort in the US for the past several years through the well-established USDOE/JAERI Collaborative Program on Fusion Neutronics. Significant contributions have been made in providing solid bases for advancing the neutronics testing capabilities in fusion reactors. This resulted from the hands-on experience gained from conducting several fusion integral experiments to quantify the prediction uncertainties of key blanket design parameters such as tritium production rate, activation, and nuclear heating, and when possible, to narrow the gap between calculational results and measurements through improving nuclear data base and codes capabilities. The current focus is to conduct the experiments in an annular configuration where the test assembly totally surrounds a simulated line source. The simulated line source is the first-of-a-kind in the scope of fusion integral experiments and presents a significant contribution to the world of fusion neutronics. The experiments proceeded through Phase IIIA to Phase IIIC in these line source simulation experiments started in 1989

  17. 233U breeding and neutron multiplying blankets for fusion reactors

    International Nuclear Information System (INIS)

    Cook, A.G.; Maniscalco, J.A.

    1975-01-01

    In this work, along with a previous paper three possible uses of 14-MeV deuterium--tritium fusion neutrons are investigated: energy production, neutron multiplication, and fissile-fuel breeding. The results presented include neutronic studies of fissioning and nonfissioning thorium systems, tritium breeding systems, various fuel options (UO 2 , UC, UC 2 , etc.), and uranium as well as refractory metal first-wall neutron-multiplying regions. A brief energy balance and an estimate of potential revenues for fusion devices are given to help illustrate the potentials of these designs

  18. Engineering the fusion reactor first wall

    International Nuclear Information System (INIS)

    Wurden, Glen; Scott, Willms

    2008-01-01

    Recently the National Academy of Engineering published a set of Grand Challenges in Engineering in which the second item listed was entitled 'Provide energy from fusion'. Clearly a key component of this challenge is the science and technology associated with creating and maintaining burning plasmas. This is being vigorously addressed with both magnetic and inertial approaches with various experiments such as ITER and NIF. Considerably less attention is being given to another key component of this challenge, namely engineering the first wall that will contain the burning plasma. This is a daunting problem requiring technologies and materials that can not only survive, but also perform multiple essential functions in this extreme environment. These functions are (1) shield the remainder of the device from radiation. (2) convert of neutron energy to useful heat and (3) breed and extract tritium to maintain the reactor fuel supply. The first wall must not contaminate the plasma with impurities. It must be infused with cooling to maintain acceptable temperatures on plasma facing and structural components. It must not degrade. It must avoid excessive build-up of tritium on surfaces, and, if surface deposits do form, must be receptive to cleaning techniques. All these functions and constraints must be met while being subjected to nuclear and thermal radiation, particle bombardment, high magnetic fields, thermal cycling and occasional impingement of plasma on the surface. And, operating in a nuclear environment, the first wall must be fully maintainable by remotely-operated manipulators. Elements of the first wall challenge have been studied since the 1970' s both in the US and internationally. Considerable foundational work has been performed on plasma facing materials and breeding blanket/shield modules. Work has included neutronics, materials fabrication and joining, fluid flow, tritium breeding, tritium recovery and containment, energy conversion, materials damage and

  19. Conceptual design of neutron diagnostic systems for fusion experimental reactor

    International Nuclear Information System (INIS)

    Iguchi, T.; Kaneko, J.; Nakazawa, M.

    1994-01-01

    Neutron measurement in fusion experimental reactors is very important for burning plasma diagnostics and control, monitoring of irradiation effects on device components, neutron source characterization for in-situ engineering tests, etc. A conceptual design of neutron diagnostic systems for an ITER-like fusion experimental reactor has been made, which consists of a neutron yield monitor, a neutron emission profile monitor and a 14-MeV spectrometer. Each of them is based on a unique idea to meet the required performances for full power conditions assumed at ITER operation. Micro-fission chambers of 235 U (and 238 U) placed at several poloidal angles near the first wall are adopted as a promising neutron yield monitor. A collimated long counter system using a 235 U fission chamber and graphite neutron moderators is also proposed to improve the calibration accuracy of absolute neutron yield determination

  20. Integrity of the first wall in fusion reactors

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    2004-07-01

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

  1. Liquid first walls for magnetic fusion energy

    International Nuclear Information System (INIS)

    Moir, R.W.

    1996-01-01

    Liquids (∼7 neutron mean free paths thick) with certain restrictions can probably be used in magnetic fusion designs between the burning plasma and the structural materials of the plant. If this works there are a number of profound advantages: lower the cost of electricity by more than 35%; remove the need to develop first wall materials saving over 4B$ in development costs; reduce the amount and kind of wastes generated in the plant; and permit a wider choice of materials. Evaporated liquid must be efficiently ionized in an edge plasma to prevent penetrating into the burning plasma and diminishing the burn rate. The fraction of evaporated material ionized is estimated to be 0.993 for Li, 0.98 for Flibe and 0.9999 for Li 17 Pb 83 . This ionized vapor would be swept along open field lines into a remote burial chamber. The most practical systems would be those with topological open field lines on the outer surface as is the case of a field reversed configuration (FRC), a Spheromak, a Z-pinch, or a mirror machine. In a Tokamak, including the Spherical Tokamak, the field lines outside the separatrix are restricted to a small volume inside the toroidal coil making for difficulties in introducing the liquid and removing the ionized vapor

  2. Material options for a commercial fusion reactor first wall

    International Nuclear Information System (INIS)

    Dabiri, A.E.

    1986-05-01

    A study has been conducted to evaluate the potential of various materials for use as first walls in high-power-density commercial fusion reactors. Operating limits for each material were obtained based on a number of criteria, including maximum allowable structural temperatures, critical heat flux, ultimate tensile strength, and design-allowable stress. The results with water as a coolant indicate that a modified alloy similar to HT-9 may be a suitable candidate for low- and medium-power-density reactor first walls with neutron loads of up to 6 MW/m 2 . A vanadium or copper alloy must be used for high-power-density reactors. The neutron wall load limit for vanadium alloys is about 14 MW 2 , provided a suitable coating material is chosen. The extremely limited data base for radiation effects hinders any quantitative assessment of the limits for copper alloys

  3. Poloidal variations in toroidal fusion reactor wall power loadings

    International Nuclear Information System (INIS)

    Carroll, M.C.; Miley, G.H.

    1985-01-01

    A geometric formulation is developed by the authors for determining poloidal variations in bremmstrahlung, cyclotron radiation, and neutron wall power loadings in toroidal fusion devices. Assuming toroidal symmetry and utilizing a numerical model which partitions the plasma into small cells, it was generally found that power loadings are highest on the outer surface of the torus, although variations are not as large as some have predicted. Results are presented for various plasma power generation configurations, plasma volume fractions, and toroidal aspect ratios, and include plasma and wall blockage effects

  4. LiWall Fusion - The New Concept of Magnetic Fusion

    International Nuclear Information System (INIS)

    Zakharov, L.E.

    2011-01-01

    Utilization of the outstanding abilities of a liquid lithium layer in pumping hydrogen isotopes leads to a new approach to magnetic fusion, called the LiWall Fusion. It relies on innovative plasma regimes with low edge density and high temperature. The approach combines fueling the plasma by neutral injection beams with the best possible elimination of outside neutral gas sources, which cools down the plasma edge. Prevention of cooling the plasma edge suppresses the dominant, temperature gradient related turbulence in the core. Such an approach is much more suitable for controlled fusion than the present practice, relying on high heating power for compensating essentially unlimited turbulent energy losses.

  5. New neutron cross sections for fusion materials studies

    International Nuclear Information System (INIS)

    Greenwood, L.R.; Smither, R.K.

    1985-01-01

    Neutron cross sections are being developed for a variety of fusion-related applications including neutron dosimetry, fusion plasma diagnostics, the activation of very long-lived isotopes, and high-energy accelerator neutron sources

  6. The fusion reactor wall is getting hot. A challenge towards the future for numerical modelling (4). Chap. 4. What is really happening in the wall?

    International Nuclear Information System (INIS)

    Murata, Isao; Konno, Chikara

    2008-01-01

    In fusion plasmas, a lot of fast neutrons with a kinetic energy of 14 MeV are generated through D-T fusion reactions. These neutrons travel deep into the first wall and are absorbed in the blanket through nuclear reactions. In the present chapter, the authors discuss what happens in the blanket with the help of computerized simulation. (T.I.)

  7. Neutron measurements in search of cold fusion

    International Nuclear Information System (INIS)

    Anderson, R.E.; Goulding, C.A.; Johnson, M.W.; Butterfield, K.B.; Gottesfeld, S.; Baker, D.A.; Springer, T.E.; Garzon, F.H.; Bolton, R.D.; Leonard, E.M.; Chancellor, T.

    1990-01-01

    We have conducted a research for neutron emission from cold fusion systems of the electrochemical type and, to a lesser extent, the high-pressure gas cell type. Using a high-efficiency well counter and an NE 213 scintillator, the experiments were conducted on the earth's surface and in a shielded cave approximately 50 ft underground. After approximately 6500 h of counting time, we have obtained no evidence for cold fusion processes leading to neutron production. However, we have observed all three types of neutron data that have been presented as evidence for cold fusion: large positive fluctuations in the neutron counting rate, weak peaks near 2.5 MeV in the neutron energy spectrum, and bursts of up to 145 neutrons in 500-μs intervals. The data were obtained under circumstances that clearly show our results to be data encountered as a part of naturally occurring neutron background, which is due primarily to cosmic rays. Thus, observing these types of data does not, of itself, provide evidence for the existence of cold fusion processes. Artifacts in the data that were due to counter misbehavior were also to lead to long-term ''neutron bursts'' whose time duration varied from several hours to several days. We conclude that any experiments which attempt to observe neutron emission must include strong steps to ensure that the experiments deal adequately with both cosmic-ray processes and counter misbehavior. 13 refs., 14 figs

  8. Neutronics of Laser Fission-Fusion Systems

    International Nuclear Information System (INIS)

    Velarde, G.

    1976-01-01

    Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-∞). (Author) 14 refs

  9. Neutronics of Laser Fission-Fusion Systems

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G

    1976-07-01

    Neutronics of Fission-Fusion microsystems inertially confined by Lasers are analysed by transport calculation, both stationary (DTF, TIHOC) and time dependent (TDA, TIHEX), discussing the results obtained for the basic parameters of the fission process (multiplication factor, neutron generation time and Rossi-{infinity}). (Author) 14 refs.

  10. Development of a Portable Fusion Neutron Generator

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Byung-Hoon; In, Sang-Ryul; Jin, Jeong-Tae; Chang, Dae-Sik; Jang, Doh-Yun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Ho [Hanyang Univ., Seoul (Korea, Republic of)

    2015-05-15

    For this purpose commercial ones, fast neutron yield from 10{sup 7} to 10{sup 11}, are supplied by several companies and research groups around the world. But internally the following limits make it difficult to develop the related application systems by domestic companies and/or research groups. - Limited life time - High price - Frequent trouble Not only to remove these limits but also to find out new internal application fields, it is necessary to develop our own domestic neutron generators. With the related technologies earned during fusion related researches, we did start to develop movable neutron generators from small one to big one, which could cover different fusion neutron yields. In this presentation the design and initial experimental results on the developed small neutron generator with a final target of 10{sup 8} n/s of 14 MeV neutrons, will be summarized.

  11. A neutron calorimeter as a fusion diagnostic

    International Nuclear Information System (INIS)

    Proctor, A.E.; Harker, Y.D.; Neischmidt, E.B.

    1986-01-01

    A calorimeter is described which is applicable as a fusion neutron diagnostic. The advantages of the device are discussed, including: low sensitivity to thermal neutrons, no heat loss to surroundings, large dynamic range, small mass resulting in fair time resolution, and small physical size. The heat generation is provided by neutron induced fissions in a foil of 238 U and a calorimeter is isothermal. The effects, advantages and disadvantages of other target materials are discussed. Also discussed are time resolution and calibration

  12. An advanced fusion neutron source facility

    International Nuclear Information System (INIS)

    Smith, D.L.

    1992-01-01

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

  13. Neutron personnel dosimetry considerations for fusion reactors

    International Nuclear Information System (INIS)

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

    1979-07-01

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

  14. Maximum neutron yeidls in experimental fusion devices

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1979-02-01

    The optimal performances of 12 types of fusion devices are compared with regard to neutron production rate, neutrons per pulse, and fusion energy multiplication, Q/sub p/ (converted to the equivalent value in D-T operation). The record values in all categories are held by the beam-injected tokamak plasma, followed by other beam-target systems. The achieved values of Q/sub p/ for nearly all laboratory plasma fusion devices (magnetically or inertially confined) are found to roughly satisfy a common empirical scaling, Q/sub p/ approx. 10 -6 E/sub in//sup 3/2/, where E/sub in/ is the energy (in kilojoules) injected into the plasma during one or two energy confinement times, or the total energy delivered to the target for inertially confined systems. Fusion energy break-even (Q/sub p/ = 1) in any system apparently requires E/sub in/ approx. 10,000 kJ

  15. Sonoluminescence: an IRaser creating cold fusion neutrons?

    International Nuclear Information System (INIS)

    Prevenslik, T.V.

    1996-01-01

    Sonoluminescence can be explained by treating the bubbles as IRasers with standing waves in resonance with the bubble dimensions. Since the IRaser resonant radiation is required to satisfy wave boundary conditions, the water molecules lining the bubble walls undergo a continuous population inversion as the bubble collapses. By stimulated processes, the Planck energy accumulates as the K b T energy of radiation photons is pumped from the surroundings through the rotational state of the water molecule. Bubble collapse occurs almost isothermally with the high IR absorptivity of the water molecule permitting the Planck energy to accumulate to 2∼6 eV only to be released by VIS-UV photon emission because of the low absorptivity of water at VIS-UV frequencies. As the IRaser cavity dimensions collapse to the spacing between water molecules at liquid density, soft x-rays at about 2 keV are predicted. But, this is less than 10 keV necessary for cold fusion so that no neutrons is directly expected yet. Therefore, it is suggested that UV laser enhancement is used to accumulate further bubble collapse energy

  16. US/Japan collaborative program on fusion reactor materials: Summary of the tenth DOE/JAERI Annex I technical progress meeting on neutron irradiation effects in first wall and blanket structural materials

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1989-01-01

    This meeting was held at Oak Ridge National Laboratory on March 17, 1989, to review the technical progress on the collaborative DOE/JAERI program on fusion reactor materials. The purpose of the program is to determine the effects of neutron irradiation on the mechanical behavior and dimensional stability of US and Japanese austenitic stainless steels. Phase I of the program focused on the effects of high concentrations of helium on the tensile, fatigue, and swelling properties of both US and Japanese alloys. In Phase II of the program, spectral and isotropic tailoring techniques are fully utilized to reproduce the helium:dpa ratio typical of the fusion environment. The Phase II program hinges on a restart of the High Flux Isotope Reactor by mid-1989. Eight target position capsules and two RB* position capsules have been assembled. The target capsule experiments will address issues relating to the performance of austenitic steels at high damage levels including an assessment of the performance of a variety of weld materials. The RB* capsules will provide a unique and important set of data on the behavior of austenitic steels irradiated under conditions which reproduce the damage rate, dose, temperature, and helium generation rate expected in the first wall and blanket structure of the International Thermonuclear Experimental Reactor

  17. The NSCL neutron wall facility

    International Nuclear Information System (INIS)

    Zecher, P.; Galonsky, A.; Kruse, J.

    1995-01-01

    The authors have constructed and installed a large-area, high-efficiency neutron detector at the NSCL. The motivation behind the detector's design was provided by their desire to improve a previous experiment, where they measured the soft-dipole-resonance parameters and ground state n-n correlations in 11 Li, and to perform similar experiments on other neutron-rich halo nuclei. The detector consists of two planes of liquid scintillator, each 4 square meters in area; it is position sensitive and is capable of neutron-γ-ray discrimination. A general overview of the detector's design and measurements of its performance in test experiments will be presented

  18. First wall response to energy disposition in conceptual laser fusion reactors

    International Nuclear Information System (INIS)

    Hovingh, J.

    1976-02-01

    Discussed are energy depositions in the first wall of various proposed laser-fusion reactors and the effect of pulse time on the stress and temperature in the first wall. Simple models can be used to estimate the temperature and stress rise from x-rays and neutrons. More complex analysis is needed to estimate the response of the first wall to reflected laser light and the pellet debris

  19. Neutron calorimeter as a fusion diagnostic

    International Nuclear Information System (INIS)

    Proctor, A.E.; Nieschmidt, E.B.

    1986-01-01

    A calorimeter is described which is applicable as a fusion neutron diagnostic. The device has the following distinct advantages: low sensitivity to thermal neutrons, large dynamic range, small mass resulting in fair time resolution, small physical size, independent calibration, little shielding required, no heat loss to surroundings, and low cost. The heat generation is provided by neutron induced fissions in a foil of 235 U or 238 U. The effects, advantages, and disadvantages of these target materials are discussed. The expected time resolution and dynamic range are estimated for both target materials

  20. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    Carvalho, S.H. de.

    1980-03-01

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

  1. Early fusion reactor neutronic calculations: A reevaluation

    International Nuclear Information System (INIS)

    Perry, R.T.

    1996-01-01

    Several fusion power plant design studies were made at a number of universities and laboratories in the late 1960s and early 1970s. These studies included such designs as the Princeton Plasma Physics Laboratory Fusion Power Plan and the University of Wisconsin UWMAK-I Reactor Neutronic analyses of the blankets and shields were part of the studies. During this time there were dissertations written on neutronic analysis systems and the results of neutronic analysis on several blanket and shield designs. The results were presented in the literature. Now in the fifth decade of fusion research, investigators often return to the earlier analyses for the neutronic results that are applicable to current blanket and shield designs, with the idea of using the older work as a basis for the new. However, the analyses of the past were made with cross-section data sets that have long been replaced with more modern versions. In addition, approximations were often made to the cross sections used because more exact data were not available. Because these results are used as guides, it is important to know if they are reproducible using more modern data. In this paper, several of the neutronic calculations made in the early studies are repeated using the MATXS-11 data library. This library is the ENDF/B-VI version of the MATXS-5 library. The library has 80 neutron groups. Tritium breeding ratios, heating rates, and fluxes are calculated and compared. This transport code used here is the one- dimensional S n code, ONEDANT. It is important to note that the calculations here are not to be considered as benchmarks because parameter and sensitivity studies were not made. They are used only to see if the results of older calculations are in reasonable agreement with a more modern library

  2. Review of the works on plasma-wall interactions in fusion reactors, 1

    International Nuclear Information System (INIS)

    Sone, Kazuho

    1975-09-01

    A review is made of the works on sputtering as one of the plasma-wall interactions in thermonuclear fusion devices. The present status and future problems are described mainly in experiments of low-energy light ions such as H + , H 2 + , D + , D 2 + and He + , heavy ions including self-ions, and fast neutrons for polycrystalline metal targets. (auth.)

  3. Advanced neutron diagnostics for ITER fusion experiments

    International Nuclear Information System (INIS)

    Kaellne, J.; Giacomelli, L.; Hjalmarsson, A.; Conroy, S.; Ericsson, G.; Johnson, M.G.; Glasser, W.; Henriksson, H.; Ronchi, E.; Sjoestrand, H.; Andersson, E.S.; Thun, J.; Weiszflog, M.; Gorini, G.; Tardocchi, M.; Popovichev, S.; Sousa, J.

    2005-01-01

    Results are presented from the neutron emission spectroscopy (NES) diagnosis of JET plasma performed with the MPR during the DTE1 campaign of 1997 and the recent TTE of 2003. The NES diagnostic capabilities at JET are presently being drastically enhanced by an upgrade of the MPR (MPRu) and a new 2.5-MeV TOF neutron spectrometer (TOFOR). The principles of MPRu and TOFOR are described and illustrated with the diagnostic role they will play in the high performance fusion experiments in the forward program of JET largely aimed at supporting ITER. The importance for the JET NES effort for ITER is discussed. (author)

  4. Neutronic assessment of strontium-90 transmutation in fusion reactors

    International Nuclear Information System (INIS)

    Parish, T.A.

    1979-01-01

    Transmutation of 90 Sr using fusion neutrons has been suggested as a possible technique for disposing of this waste nuclide. For transmutation to be attractive, high transmutation rates relative to natural decay are required. Effective half-lives for 90 Sr were computed for fusion reactor blankets constructed of various materials. To obtain satisfactory transmutation rates, fusion reactors with high first wall neutron currents and with highly moderating blankets were found to be necessary. An effective half-life for 90 Sr of 90 Sr inventory and the number of burners required for various fission usage scenarios. Efficient and fast chemical separations were needed to reap the benefits of a short effective half-life. For the fusion burners considered, it was found that the 90 Sr inventory could not be reduced to less than one-fourth of the inventory without transmutation if fission usage continued at a constant rate. Such a reduction is not sufficient to justify the transmutation disposal of 90 Sr

  5. Irradiation damage of ferritic/martensitic steels: Fusion program data applied to a spallation neutron source

    International Nuclear Information System (INIS)

    Klueh, R.L.

    1997-01-01

    Ferritic/martensitic steels were chosen as candidates for future fusion power plants because of their superior swelling resistance and better thermal properties than austenitic stainless steels. For the same reasons, these steels are being considered for the target structure of a spallation neutron source, where the structural materials will experience even more extreme irradiation conditions than expected in a fusion power plant first wall (i.e., high-energy neutrons that produce large amounts of displacement damage and transmutation helium). Extensive studies on the effects of neutron irradiation on the mechanical properties of ferritic/martensitic steels indicate that the major problem involves the effect of irradiation on fracture, as determined by a Charpy impact test. There are indications that helium can affect the impact behavior. Even more helium will be produced in a spallation neutron target material than in the first wall of a fusion power plant, making helium effects a prime concern for both applications. 39 refs., 10 figs

  6. Neutron measurements as fusion plasma diagnostics

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Hoek, M.

    1993-01-01

    Neutron measurements play important roles as the diagnostics of many aspects of the plasma in large tokamak devices such as JT-60U and JET. In the d-d discharges of JT-60U, the most important application of the neutron measurement is the investigation of the fusion performance using fission chambers. The ion velocity distribution function, and the triton slowing down are investigated by the neutron spectrometer and the 14 MeV neutron detector, respectively. TANSY is a combined proton-recoil and neutron time-of flight spectrometer for 14 MeV neutrons to be used during the d-t phase at JET. The detection principle is based on the measurements of the flight time of a scattered initial neutron and the energy of a corresponding recoil proton. The scattering medium is a polyethylene foil. The resolution and efficiency, using a thin foil (0.95 mg/cm 2 ), is 155 keV and 1.4x10 -5 cm 2 , respectively. (author)

  7. Cold fusion produces more tritium than neutrons

    International Nuclear Information System (INIS)

    Rajagopalan, S.R.

    1989-01-01

    The results of the major cold fusion experiments performed in various laboratories of the world and attempts to explain them are reviewed in brief. Particular reference is made to the experiments carried out in the Bhabha Atomic Research Centre (BARC), Bombay. In BARC experiments, it is found that tritium is the primary product of cold fusion. Author has put forward two hypothetical pictures of D-D fusion. (1) When a metal like Pd or Ti is loaded with D 2 , a crack forms. Propogation of such a crack accelerates deuterons which bombard Pd D 2 /D held by Pd or Ti leading to neutron capture or tritium formation with the release of protons and energy. The released protons might transfer its energy to some other deuteron and a chain reaction is started. This chain reaction terminates when a substantial portion of D in the crack tip is transmuted. This picture explains fusion reaction bursts and the random distribution of reaction sites, but does not explain neutron emission. (2) The deuterons accelerated by a propogating crack may hit a Pd/Ti nucleus instead of a deuterium nucleus and may transmute Pd/Ti. (M.G.B.). 18 refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-10-01

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

  9. The reversed-field-pinch (RFP) fusion neutron source: A conceptual design

    International Nuclear Information System (INIS)

    Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Werley, K.A.

    1989-01-01

    The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at ∼5-MW/m 2 steady-state DT fusion neutron wall loading and ∼124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low input power (∼206 MW) source of DT neutrons for large-volume (∼10 m 3 ), high-fluence (3.4 MW yr/m 2 ) fusion nuclear materials and technology testing. 19 refs., 15 figs., 9 tabs

  10. Neutron cross section libraries for analysis of fusion neutronics experiments

    International Nuclear Information System (INIS)

    Kosako, Kazuaki; Oyama, Yukio; Maekawa, Hiroshi; Nakamura, Tomoo

    1988-03-01

    We have prepared two computer code systems producing neutron cross section libraries to analyse fusion neutronics experiments. First system produces the neutron cross section library in ANISN format, i.e., the multi-group constants in group independent format. This library can be obtained by using the multi-group constant processing code system MACS-N and the ANISN format cross section compiling code CROKAS. Second system is for the continuous energy cross section library for the MCNP code. This library can be obtained by the nuclear data processing system NJOY which generates pointwise energy cross sections and the cross section compiling code MACROS for the MCNP library. In this report, we describe the production procedures for both types of the cross section libraries, and show six libraries with different conditions in ANISN format and a library for the MCNP code. (author)

  11. Automation of fusion first wall design using artificial intelligence technique

    International Nuclear Information System (INIS)

    Yoshimura, Shinobu; Yagawa, Genki; Mochizuki, Yoshihiko

    1990-01-01

    This paper describes the application of artificial intelligence techniques to a design automation of the fusion first wall to be operated in the complex environment where huge electromagnetic and thermal loading as well as heavy neutron irradiation occur. As a basic strategy of designing structure shape considering many coupled phenomena, an ordinary design procedure based on the generate and test strategy is adopted because of its simplicity and broad applicability. To automate the design procedure with maintaining its flexibility, extensibility and efficiency, artificial intelligence techniques are utilized in the following. An object-oriented knowledge representation technique is adopted to store knowledge modules, that is, objects, related to the first wall design, while a data-flow processing technique is utilized as an inference mechanism among the knowledge modules. These techniques realize the flexibility and extensibility of the system. Moreover, as an efficient design modification mechanism, which is essential in a design process, an empirical approach based on experts' empirical knowledge and a mathematical approach based on a kind of numerical sensitivity analysis are introduced. The developed system is applied to a simple example of the design of a two-dimensional model of the first wall with a cooling channel, and its fundamental performance is clearly demonstrated. (author)

  12. Fusion enhancement in the reactions of neutron-rich nuclei

    International Nuclear Information System (INIS)

    Bian Baoan; Zhang Fengshou; Zhou Hongyu

    2009-01-01

    The neutron-rich fusion reactions are investigated systematically using the improved isospin dependent quantum molecular dynamics model. By studying the systematic dependence of fusion barrier on neuron excess, we find the enhancement of the fusion cross sections for neutron-rich nuclear reactions that give the lowered static Coulomb barriers. The calculated fusion cross sections agree quantitatively with the experimental data. We further discuss the mechanism of the fusion enhancement of the cross sections for neutron-rich nuclear reactions by analyzing the dynamical lowering of the Coulomb barrier that is attributed to the enhancement of the N/Z ratio at the neck region.

  13. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1983-06-01

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

  14. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    Barre, F.

    1984-02-01

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

  15. Fuel cycle for a fusion neutron source

    Science.gov (United States)

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

    2015-12-01

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

  16. ANITA-2000, Isotope Inventories from Neutron Irradiation, for Fusion Applications

    International Nuclear Information System (INIS)

    Cepraga, Dan-Gabriel

    2000-01-01

    1 - Description of program or function: ANITA-2000 is a code package for the activation characterisation of materials exposed to neutrons in fusion machines. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation, continuous or stepwise. It provides activity, atomic density, decay heat, biological hazard, clearance index and gamma-ray source spectra at shut down and for different cooling times. An interactive utility module, MODBIN, to produce the neutron activation cross sections libraries in the required binary ANITA-4M Format, is also included. The GRANITA interactive module may plot activation parameters as a function of the cooling time. The main improvements include: -the number of irradiation time intervals has been increased to 2000; -different neutron wall loading can be used for each burn time interval; -the photon source calculation in the 18 energy group Scale structure has been added; -the clearance index can be calculated. In addition the code language has been standardized to Fortran '95 - by maintaining the backward compatibility (except for the time/date routines) - so as the same code package can be compiled and run on Unix environment and on PC, both under DOS-Windows and under Linux. 2 - Methods: The mathematical solution of the problem is given in analytical form using recurrence relations. Double precision arithmetic is used

  17. JAERI/U.S. collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Mori, Takamasa; Kosako, Kazuaki; Oyama, Yukio; Nakamura, Tomoo

    1989-10-01

    Phase IIa and IIb experiments of JAERI/U.S. Collaborative Program on Fusion Blanket Neutronics have been performed using the FNS facility at JAERI. The phase IIa experimental systems consist of the Li 2 O test region, the rotating neutron target and the Li 2 CO 3 container. In phase IIb, a beryllium layer is added to the inner wall to investigate a multiplier effect. Measured parameters are source characteristics by a foil activation method and spectrum measurements using both NE-213 and proton recoil counters. The measurements inside the Li 2 O region included tritium production rates, reaction rate by foil activation and neutron spectrum measurements. Analysis for these parameters was performed by using two dimensional discrete ordinate codes DOT3.5 and DOT-DD, and a Monte Carlo code MORSE-DD. The nuclear data used were based on JENDL3/PR1 and PR2. ENDF/B-IV, V and the FNS file were used as activation cross sections. The configurations analysed for the test region were a reference, a beryllium front and a beryllium sandwiched systems in phase IIa, and a reference and a beryllium front with first wall systems in phase IIb. This document describes the results of analysis and comparison between the calculations and the measurements. The prediction accuracy of key parameters in a fusion reactor blanket are examined. The tritium production rates can be well predicted in the reference systems but are fairly underestimated in the system with a beryllium multiplier. Details of experiments and the experimental techniques are described separately in the another report. (author)

  18. Collection of experimental data for fusion neutronics benchmark

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Yamamoto, Junji; Ichihara, Chihiro; Ueki, Kotaro; Ikeda, Yujiro.

    1994-02-01

    During the recent ten years or more, many benchmark experiments for fusion neutronics have been carried out at two principal D-T neutron sources, FNS at JAERI and OKTAVIAN at Osaka University, and precious experimental data have been accumulated. Under an activity of Fusion Reactor Physics Subcommittee of Reactor Physics Committee, these experimental data are compiled in this report. (author)

  19. Neutron detector for fusion reaction-rate measurements

    International Nuclear Information System (INIS)

    Lerche, R.A.; Phillion, D.W.; Tietbohl, G.L.

    1993-01-01

    We have developed a fast, sensitive neutron detector for recording the fusion reaction-rate history of inertial-confinement fusion (ICF) experiments. The detector is based on the fast rise-time of a commercial plastic scintillator (BC-422) and has a response 7 neutrons

  20. Fuel cycle for a fusion neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Ananyev, S. S., E-mail: Ananyev-SS@nrcki.ru; Spitsyn, A. V., E-mail: spitsyn-av@nrcki.ru; Kuteev, B. V., E-mail: Kuteev-BV@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m{sup 3}Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

  1. Laser Intertial Fusion Energy: Neutronic Design Aspects of a Hybrid Fusion-Fission Nuclear Energy System

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, Kevin James [Univ. of California, Berkeley, CA (United States)

    2010-04-08

    This study investigates the neutronics design aspects of a hybrid fusion-fission energy system called the Laser Fusion-Fission Hybrid (LFFH). A LFFH combines current Laser Inertial Confinement fusion technology with that of advanced fission reactor technology to produce a system that eliminates many of the negative aspects of pure fusion or pure fission systems. When examining the LFFH energy mission, a significant portion of the United States and world energy production could be supplied by LFFH plants. The LFFH engine described utilizes a central fusion chamber surrounded by multiple layers of multiplying and moderating media. These layers, or blankets, include coolant plenums, a beryllium (Be) multiplier layer, a fertile fission blanket and a graphite-pebble reflector. Each layer is separated by perforated oxide dispersion strengthened (ODS) ferritic steel walls. The central fusion chamber is surrounded by an ODS ferritic steel first wall. The first wall is coated with 250-500 μm of tungsten to mitigate x-ray damage. The first wall is cooled by Li17Pb83 eutectic, chosen for its neutron multiplication and good heat transfer properties. The Li17Pb83 flows in a jacket around the first wall to an extraction plenum. The main coolant injection plenum is immediately behind the Li17Pb83, separated from the Li17Pb83 by a solid ODS wall. This main system coolant is the molten salt flibe (2LiF-BeF2), chosen for beneficial neutronics and heat transfer properties. The use of flibe enables both fusion fuel production (tritium) and neutron moderation and multiplication for the fission blanket. A Be pebble (1 cm diameter) multiplier layer surrounds the coolant injection plenum and the coolant flows radially through perforated walls across the bed. Outside the Be layer, a fission fuel layer comprised of depleted uranium contained in Tristructural-isotropic (TRISO) fuel particles

  2. Conventional sources of fast neutrons in 'cold fusion' experiments

    International Nuclear Information System (INIS)

    Cribier, M.; Spiro, M.; Favier, J.

    1989-04-01

    In 'cold fusion' experiments with heavy water a source of neutrons is the dissociation of deuterium induced by alpha particles emitted by natural occurring radioisotopes. We evaluate the rate of fast neutron emission as a function of the concentration of U, Th, Rn in contact with deuterium and discuss the possibility that the neutrons claimed to have been observed in 'cold fusion' experiments could be due to this conventional source

  3. Development of vanadium base alloys for fusion first-wall/blanket applications

    International Nuclear Information System (INIS)

    Smith, D.L.; Chung, H.M.; Loomis, B.A.; Matsui, H.; Votinov, S.; VanWitzenburg, W.

    1994-01-01

    Vanadium alloys have been identified as a leading candidate material for fusion first-wall/blanket applications. Certain vanadium alloys exhibit favorable safety and environmental characteristics, good fabricability, high temperature and heat load capability, good compatibility with liquid metals and resistance to irradiation damage effects. The current focus is on vanadium alloys with (3-5)% Cr and (3-5)% Ti with a V-4Cr-4Ti alloy as the leading candidate. Preliminary results indicate that the crack-growth rates of certain alloys are not highly sensitive to irradiation. Results from the Dynamic Helium Charging Experiment (DHCE) which simulates fusion relevant helium/dpa ratios are similar to results from neutron irradiated material. This paper presents an overview of the recent results on the development of vanadium alloys for fusion first wall/blanket applications

  4. Neutronic analysis of fusion tokamak devices by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

    2011-01-01

    A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

  5. Performance limits for fusion first-wall structural materials

    International Nuclear Information System (INIS)

    Smith, D.L.; Majumdar, S.; Billone, M.; Mattas, R.

    2000-01-01

    Key features of fusion energy relate primarily to potential advantages associated with safety and environmental considerations and the near endless supply of fuel. However, high-performance fusion power systems will be required in order to be an economically competitive energy option. As in most energy systems, the operating limits of structural materials pose a primary constraint to the performance of fusion power systems. In the case of fusion power, the first-wall/blanket system will have a dominant impact on both economic and safety/environmental attractiveness. This paper presents an assessment of the influence of key candidate structural material properties on performance limits for fusion first-wall blanket applications. Key issues associated with interactions of the structural materials with the candidate coolant/breeder materials are discussed

  6. Heat transfer models for fusion blanket first walls

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1977-01-01

    In the development of magnetically confined fusion reactors, the ability to cool the first wall, i.e., the first material surface interfacing the plasma, appears to be a critical factor involved in establishing the wall load limit. In order to understand the thermal behavior of the first wall time-dependent, one-dimensional heat conduction models are reviewed with differing modes of heat extraction and cooling

  7. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  8. Beam plasma 14 MeV neutron source for fusion materials development

    International Nuclear Information System (INIS)

    Ravenscroft, D.; Bulmer, D.; Coensgen, F.; Doggett, J.; Molvik, A.; Souza, P.; Summers, L.; Williamson, V.

    1991-09-01

    The conceptual engineering design and expected performance for a 14 MeV DT neutron source is detailed. The source would provide an intense neutron flux for accelerated testing of fusion reactor materials. The 150-keV neutral beams inject energetic deuterium atoms, that ionize, are trapped, then react with a warm (200 eV), dense tritium target plasma. This produces a neutron source strength of 3.6 x 10 17 n/sec for a neutron power density at the plasma edge of 5--10 MW/m 2 . This is several times the ∼2 MW/m 2 anticipated at the first wall of fusion reactors. This high flux provides accelerated end-of-life tests of 1- to 2-year duration, thus making materials development possible. The modular design of the source and the facilities are described

  9. Fusion neutron detector calibration using a table-top laser generated plasma neutron source

    International Nuclear Information System (INIS)

    Hartke, R.; Symes, D.R.; Buersgens, F.; Ruggles, L.E.; Porter, J.L.; Ditmire, T.

    2005-01-01

    Using a high intensity, femtosecond laser driven neutron source, a high-sensitivity neutron detector was calibrated. This detector is designed for observing fusion neutrons at the Z accelerator in Sandia National Laboratories. Nuclear fusion from laser driven deuterium cluster explosions was used to generate a clean source of nearly monoenergetic 2.45 MeV neutrons at a well-defined time. This source can run at 10 Hz and was used to build up a clean pulse-height spectrum on scintillating neutron detectors giving a very accurate calibration for neutron yields at 2.45 MeV

  10. Development of fusion first-wall radiation damage facilities

    International Nuclear Information System (INIS)

    McElroy, R.J.; Atkins, T.

    1986-11-01

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

  11. Neutronics issues in fusion-fission hybrid reactor design

    International Nuclear Information System (INIS)

    Liu Chengan

    1995-01-01

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

  12. Graphs of neutron cross section data for fusion reactor development

    International Nuclear Information System (INIS)

    Asami, Tetsuo; Tanaka, Shigeya

    1979-03-01

    Graphs of neutron cross section data relevant to fusion reactor development are presented. Nuclides and reaction types in the present compilation are based on a WRENDA request list from Japan for fusion reactor development. The compilation contains various partial cross sections for 55 nuclides from 6 Li to 237 Np in the energy range up to 20 MeV. (author)

  13. Extension of the AUS reactor neutronics system for application to fusion blanket neutronics

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1984-03-01

    The AUS modular code scheme for reactor neutronics computations has been extended to apply to fusion blanket neutronics. A new group cross-section library with 200 neutron groups, 37 photon groups and kerma factor data has been generated from ENDF/B-IV. The library includes neutron resonance subgroup parameters and temperature-dependent data for thermal neutron scattering matrices. The validity of the overall calculation system for fusion applications has been checked by comparison with a number of published conceptual design studies

  14. The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

    The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

  15. Intense neutron source facility for the fusion energy program

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

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

  17. Disposal of activated fusion wall materials

    International Nuclear Information System (INIS)

    Blink, J.A.; Dorn, D.W.; Maninger, R.C.

    1983-08-01

    We have used NRC's low-level waste disposal regulation (10CFR61) to classify activated fusion reactor structural materials. The limits set by the NRC in 10CFR61 will require extremely expensive steels with degraded properties, even when the limits are adjusted to give credit for use of an expensive hot waste disposal facility. Both the expense and the poorer properties could have a negative impact on reactor safety, thus subverting the overall goals of the NRC family of regulations. Following this initial study, we have examined the methodology used by the NRC to set waste concentration limits. For a long-lived gamma emitter like 94 Nb, direct gamma dose to an intruding home builder dominates the limit setting process. Of all the tests applied to the waste, the controlling test which sets the lowest limit ignores all the engineered intrusion barriers which are themselves required by the same regulation. If even a small fraction of the barriers remain intact (an extremely likely event), the 94 Nb limit could be increased from the 0.2 Ci/m 3 in 10CFR61 to 1100 Ci/m 3 without exceeding the limits set for personnel exposure. Similarly, cautious application of the 10CFR61 methodology to other radioisotopes of interest to fusion designers will result in limits which are more in line with the unique nature of fusion energy

  18. Shock-tube study of fusion plasma-wall interactions

    International Nuclear Information System (INIS)

    Gross, R.A.; Tien, J.K.; Jensen, B.; Panayotou, N.F.; Feinberg, B.

    1977-01-01

    Theoretical and experimental studies have been made of phenomena which occur when a hot (T 1 approximately equal to 6 x 10 6 0 K), dense (n approximately equal to 10 16 cm -3 ), deuterium plasma containing a transverse magnetic field is brought into sudden contact with a cold metal wall. These studies are motivated by the need to understand plasma and metallurgical conditions at the first-wall of a fusion reactor. Experiments were carried out in the Columbia high energy electromagnetic shock tube. Computational simulation was used to investigate the detailed physics of the fusion plasma boundary layer which develops at the wall. The rate of energy transfer from the plasma to the wall was calculated and conditions under which surface melting occurs are estimated. Experimental measurements of plasma-wall heat transfer rates up to 3 x 10 5 watts/cm 2 were obtained and agreement with computed values are good. Fusion reactor first-wall materials have been exposed to 6.0 x 10 21 eV cm -2 (1,000 shots) of deuterium plasma bombardment. Scanning electron micrograph photographs show preferential erosion at grain boundaries, formation of deuterium surface blisters, and evidence of local surface melting. Some cracking is observed along grain boundaries, and a decrease in tensile ductiity is measured

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

    International Nuclear Information System (INIS)

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

    1980-12-01

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

  20. Lifetime analysis for fusion reactor first walls and divertor plates

    International Nuclear Information System (INIS)

    Horie, T.; Tsujimura, S.; Minato, A.; Tone, T.

    1987-01-01

    Lifetime analysis of fusion reactor first walls and divertor plates is performed by (1) a one-dimensional analytical plate model, and (2) a two-dimensional elastic-plastic finite element method. Life-limiting mechanisms and the limits of applicability for these analysis methods are examined. Structural design criteria are also discussed. (orig.)

  1. First wall costs of an ion-beam fusion reactor

    International Nuclear Information System (INIS)

    Hovingh, J.

    1977-08-01

    This paper parametrically investigates the effects of microexplosion energy on the first wall costs of a 4000 MW/sub t/ ion-beam initiated, inertially confined fusion reactor for several first wall materials. The thermodynamic models and the results for microexplosion energies between 400 and 4000 MJ are presented. A solid stainless steel or a composite isotropic graphite over stainless steel first wall can operate for a year at a cost of 0.6 mills per kWh gross electric power output

  2. Characterization of a deuterium-deuterium plasma fusion neutron generator

    Science.gov (United States)

    Lang, R. F.; Pienaar, J.; Hogenbirk, E.; Masson, D.; Nolte, R.; Zimbal, A.; Röttger, S.; Benabderrahmane, M. L.; Bruno, G.

    2018-01-01

    We characterize the neutron output of a deuterium-deuterium plasma fusion neutron generator, model 35-DD-W-S, manufactured by NSD/Gradel-Fusion. The measured energy spectrum is found to be dominated by neutron peaks at 2.2 MeV and 2.7 MeV. A detailed GEANT4 simulation accurately reproduces the measured energy spectrum and confirms our understanding of the fusion process in this generator. Additionally, a contribution of 14 . 1 MeV neutrons from deuterium-tritium fusion is found at a level of 3 . 5%, from tritium produced in previous deuterium-deuterium reactions. We have measured both the absolute neutron flux as well as its relative variation on the operational parameters of the generator. We find the flux to be proportional to voltage V 3 . 32 ± 0 . 14 and current I 0 . 97 ± 0 . 01. Further, we have measured the angular dependence of the neutron emission with respect to the polar angle. We conclude that it is well described by isotropic production of neutrons within the cathode field cage.

  3. Fusion Engineering Device (FED) first wall/shield design

    International Nuclear Information System (INIS)

    Sager, P.H.; Fuller, G.; Cramer, B.; Davisson, J.; Haines, J.; Kirchner, J.

    1981-01-01

    The torus of the Fusion Engineering Device (FED) is comprised of the bulk shield and its associated spool lstructure and support system, the first wall water-cooled panel and armor systems, and the pumped limiter. The bulk shielding is provided by ten shield sectors that are installed in the spool structure in such a way as to permit extraction of the sectors through the openings between adjacent toroidal field coils with a direct radial movement. The first wall armor is installed on the inboard and top interior walls of these sectors, and the water-cooled panels are installed on the outboard interior walls and the pumped limiter in the bottom of the sectors. The overall design of the first wall and shield system is described in this paper

  4. Liquid wall boiler and moderator (BAM) for heavy ion-pellet fusion reactors

    International Nuclear Information System (INIS)

    Powell, J.R.; Lazareth, O.; Fillo, J.

    1977-11-01

    Thick liquid wall blankets appear to be of great promise for heavy ion pellet fusion reactors. They avoid the severe problems of intense radiation and blast damage that would be encountered with solid blanket structures. The liquid wall material can be chosen so that its vapor pressure at the working temperature of the power cycle is well below the value at which it might interfere with the propagation of the heavy ion beam. The liquid wall can be arranged so that it does not contact any surrounding solid structure when the pellet explosion occurs, including the ends. The ends can be magnetically closed just before the pellet explosion, or a time phased flow can be used, which will leave a clear central zone into which the pellet is injected. Parametric analysis comparing three candidate liquid wall materials were carried out. The three materials were lithium, flibe, and lead (with a low concentration of disolved lithium). Lead appeared to be the best choice for the liquid wall, although any of the three should allow a practical reactor system. The parametric analyses examined the effects of pellet yield (0 to 10 GJ), pellet mass (3 g to 3 kg), liquid wall thickness (10 cm to 80 cm), vapor condensation time (0 to 10 milliseconds), degree of neutron moderation in the pellet (none to 100%), liquid wall chamber size (radius of 1.5 meters to 4 meters), Pb/Li 6 ratio (100 to 5,000), and thickness of graphite moderating zone behind the liquid wall

  5. Fusion reaction using low energy neutron-excess nucleus beam

    International Nuclear Information System (INIS)

    Fukuda, Tomokazu

    1994-01-01

    The present state and the plan of the experiment of measuring the fusion reaction near barriers by using neutron-excess nucleus beam, which has been advanced at RIKEN are reported. One of the purposes of this experiment is the feasibility investigation of the fusion reaction by using neutron-excess nuclei, which is indispensable for synthesizing superheavy elements. It is intended to systematically explore some enhancing mechanism in the neutron-excess nuclei which are unfavorable in beam intensity. This research can become the good means to prove the dynamic behavior of the neutrons on the surfaces of nuclei in reaction. The fusion reaction of 27 Al + Au was measured by using the stable nucleus beam of 27 Al, and the results are shown. In order to know the low energy fusion reaction of 11 Li and 11 Be which are typical halo nuclei, the identification by characteristic α ray of composite nuclei is carried out in 7,9,11 Li + 209 Bi and 9,10,11 Be + 208 Pb. A new detector having high performance, New MUSIC, is being developed. As the experiment by using this detector, the efficient measurement of the fusion reaction by using heavy neutron-excess nuclei up to Ni is considered. An example of 8 Li + α → 11 B + n reaction for celestial body physics is mentioned. (K.I.)

  6. Some stress-related issues in tokamak fusion reactor first walls

    International Nuclear Information System (INIS)

    Majumdar, S.; Pai, B.; Ryder, R.H.

    1987-01-01

    Recent design studies of a tokamak fusion power reactor and of various blankets have envisioned surface heat fluxes on the first wall ranging from 0.1 to 1.0 MW/m 2 , and end-of-life irradiation fluences ranging from 100 dpa for the austenitic stainless steels to as high as 250 dpa for postulated vanadium alloys. Some tokamak blankets, particularly those using helium or liquid metal as coolant/breeder, may have to operate at relatively high coolant pressures so that the first wall may be subjected to high primary stress in addition to high secondary stresses such as thermal stresses or stresses due to constrained swelling. The present paper focusses on the various problems that may arise in the first wall because of stress and high neutron fluence, and discusses some of the design solutions that have been proposed to overcome these problems

  7. Neutron pre-emission at the fusion of 11 Li halo nuclei with Si targets

    International Nuclear Information System (INIS)

    Petrascu, M.; Isbasescu, A.; Petrascu, H.; Bordeanu, C.; David, I.; Lazar, I.; Mihai, I.; Vaman, G.; Tanihata, I.; Kobayashi, T.; Korsheninnikov, A.; Fukuda, S.; Kumagai, H.; Momota, S.; Ozawa, A.; Yoshida, K.; Nikolski, E.; Giurgiu, M.

    1997-01-01

    In this contribution, the first experiment on fusion of 11 Li halo nuclei with Si targets is reported. A novel effect consisting of a large neutron pre-emission probability in the fusion process was observed. The neutron halo nuclei are characterized by very large matter radii, small separation energy and small internal momentum of the valence neutrons. Until now, the halo nuclei were investigated mostly by elastic, inelastic scattering and breakup processes. It was recently predicted that due to the very large dimension of 11 Li, one may expect, that in a fusion experiment on a light target, the valence neutrons will not be absorbed together with the 9 Li core, but will be emitted in the early stage of the reaction process. The experiment aiming to check this expectation, performed at RIKEN-RIPS facility, is described. In the experimental arrangement, three main parts are present: the first part contains the detectors used for the control, identification and determination of the beam characteristics; the second part consists of a Multiple Sampling Ionisation Chamber (MUSIC), used for identification of the inclusive evaporation residue spectra produced in the detector-target; the third part consists of two wall neutron detectors, each made up of 15 plastic scintillators. This detector was used for the energy and position determination of the neutrons originating from the target. The projectile energy range was 11.2 - 15.2 AMeV, centered at 13 AMeV. The neutrons resulting from the reaction were measured by time-of-light technique. The position on the 'wall' of the detected neutrons could be also determined. The measured neutron spectra from 11 Li and 9 Li are shown. A marked different between the two spectra was found and it is explained by the contribution of a large amount of pre-emission (breakup) processes, in case of 11 Li projectiles. The position spectra point out the evaporation origin of the neutrons in case of 9 Li projectiles while for 11 Li only the

  8. Ion bombardment effects on the fatigue life of stainless steel under simulated fusion first wall conditions

    International Nuclear Information System (INIS)

    Kohse, G.; Harling, O.K.

    1983-01-01

    Pressurized tube specimens have been exposed to simultaneous multi-energy surface ion bombardment, fast neutron irradiation and stress and temperature cycling, in a simulation of a possible fusion reactor first wall environment. After ion bombardments equivalent to months-years of reactor operation and up to 30,000 cycles, no detrimental effects on post-irradiation fatigue life were found. The ion damage is found to enhance surface cracking, but this effect is limited to the several micron surface layer in which the ions are implanted

  9. Determination of the Jet Neutron Rate and Fusion Power using the Magnetic Proton Recoil Neutron Spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Sjoestrand, Henrik

    2003-01-01

    In this thesis a new independent method has been developed to enable precise measurements of neutron yields and rates from fusion plasmas and thereby determining the fusion power and fusion energy. The new method, together with the associated diagnostics, can provide information of great importance to present and future high fusion yield experiments, such as the Joint European Torus (JET) tokamak and the International Thermonuclear Experiment Reactor (ITER). The method has been applied to data from high fusion rate experiments from the tritium campaign at JET. By using the count-rate from the Magnetic Proton Recoil (MPR) neutron spectrometer the number of neutrons in the spectrometer's line of sight has been calculated. To be able to do this, all relevant factors between the plasma and the instrument have been evaluated. The number of neutrons in the MPR line of sight has been related to the total number of produced neutrons in the plasma by using information on the neutron emission profile. The achieved results have been compared with other JET neutron diagnostic data and the agreement is shown to be very good.

  10. Linear induction accelerators for fusion and neutron production

    International Nuclear Information System (INIS)

    Barletta, W.A.; California Univ., Los Angeles, CA

    1993-08-01

    Linear induction accelerators (LIA) with pulsed power drives can produce high energy, intense beams or electrons, protons, or heavy ions with megawatts of average power. The continuing development of highly reliable LIA components permits the use such accelerators as cost-effective beam sources to drive fusion pellets with heavy ions, to produce intense neutron fluxes using proton beams, and to generate with electrons microwave power to drive magnetic fusion reactors and high gradient, rf-linacs

  11. 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.

  12. First wall fusion blanket temperature variation - slab geometry

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1978-01-01

    The first wall of a fusion blanket is approximated by a slab, with the surface facing the plasma subjected to an applied heat flux, while the rear surface is convectively cooled. The relevant parameters affecting the heat transfer during the early phases of heating as well as for large times are established. Analytical solutions for the temperature variation with time and space are derived. Numerical calculations for an aluminum and stainless steel slab are performed for a wall loading of 1 MW(th)/m 2 . Both helium and water cooling are considered. (Auth.)

  13. First wall lifetime of the near term fusion reactors

    International Nuclear Information System (INIS)

    Matera, R.; Botti, S.; Cerrai, G.

    1985-01-01

    A sensitivity analysis of the influence of the operating conditions and of the design parameters over the first wall lifetime was performed by means of the computer program smile. In the range of operating conditions typical of an experimental fusion reactor like NET/INTOR and for a type AISI 316 stainless steel structural material, fatigue damage and fatigue crack growth are the limiting failure mechanisms of the first wall. The analysis shows in graphical form the limits of the allowable range of operating conditions or of design parameters

  14. Physics of plasma-wall interactions in controlled fusion

    International Nuclear Information System (INIS)

    Post, D.E.; Behrisch, R.

    1984-01-01

    In the areas of plasma physics, atomic physics, surface physics, bulk material properties and fusion experiments and theory, the following topics are presented: the plasma sheath; plasma flow in the sheath and presheath of a scrape-off layer; probes for plasma edge diagnostics in magnetic confinement fusion devices; atomic and molecular collisions in the plasma boundary; physical sputtering of solids at ion bombardment; chemical sputtering and radiation enhanced sublimation of carbon; ion backscattering from solid surfaces; implantation, retention and release of hydrogen isotopes; surface erosion by electrical arcs; electron emission from solid surfaces;l properties of materials; plasma transport near material boundaries; plasma models for impurity control experiments; neutral particle transport; particle confinement and control in existing tokamaks; limiters and divertor plates; advanced limiters; divertor tokamak experiments; plasma wall interactions in heated plasmas; plasma-wall interactions in tandem mirror machines; and impurity control systems for reactor experiments

  15. First wall thermomechanical stress analysis in a fusion ignition experiment

    International Nuclear Information System (INIS)

    Rodin, G.; Carrera, R.; Howell, J.; Hwang, Y.L.; Montalvo, E.; Ordonez, C.; Dong, J.Q.

    1990-01-01

    The fusion ignition experiment IGNITEX + has been proposed as a low cost means of producing and controlling fusion ignited plasmas for scientific study. A single-turn-coil tokamak plasmas for scientific study. A single-turn-coil tokamak cryogenically precooled at liquid nitrogen temperature is used to produce 20 T fields and 12 MA plasma currents so that high-density ohmic ignition is possible. The high-field, high-density operation should maintain the plasma relatively free of wall impurities. In order to minimize plasma cooling, a low-Z first wall is considered for IGNITEX. The IGNITEX design philosophy emphasizes simplicity and low cost. A limiterless, smooth first will without files and plates is proposed. A low-Z material is applied by plasma jet techniques over a resistive vacuum vessel. This design is thought to be adequate for a magnetic fusion ignition experiment. Maintenance and operation of the first wall system is significantly simplified when compared to conventional designs

  16. Production of Medical isotope Technecium-99 from DT Fusion neutrons

    Science.gov (United States)

    Boguski, John; Gentile, Charles; Ascione, George

    2011-10-01

    High energy neutrons produced in DT fusion reactors have a secondary application for use in the synthesis of valuable man-made isotopes utilized in industry today. One such isotope is metastable Technecium-99 (Tc99m), a low energy gamma emitter used in ~ 85% of all medical imaging diagnostics. Tc99m is created through beta decay of Molybdenum-99 (Mo99), which itself has only a 66 hour half-life and must be created from a neutron capture by the widely available and stable isotope Molydenum-98. Current worldwide production of Tc99m occurs in just five locations and relies on obtaining the fission byproduct Mo99 from highly enriched Uranium reactors. A Tc99m generator using DT fusion neutrons, however, could potentially be operated at individual hospitals and medical facilities without the use of any fissile material. The neutron interaction of the DT neutrons with Molybdenum in a potential device geometry was modeled using Monte Carlo neutron transport code MCNP. Trial experiments were also performed to test the viability of using DT neutrons to create ample quantities of Tc99m. Modeling and test results will follow.

  17. Neutron multiplier alternative for fusion reactor blankets

    International Nuclear Information System (INIS)

    Taczanowski, S.

    1980-01-01

    A proposal is given to replace neutron multiplier needed to enable low lithium and tritium inventories simultaneously assuring sufficient production of tritium, by an efficient moderator ( 7 LiH or 7 LiD). The advantageous effect of the intensified neutron energy degradation is due to the 1/v character of the main tritium producing reaction. The slowing-down medium is designed to be the source of moderated neutrons for the surrounding Li ( 6 Li enriched) region where the most of tritium is to be produced. The surplus tritium production remains stored in the moderator zone. Some preliminary calculations illustrating the above concept were carried out and the neutron flux and tritium production distributions are presented. The indications regarding further studies are also suggested. (author)

  18. Neutron-transparent first wall for module testing

    International Nuclear Information System (INIS)

    Fuller, G.M.; Cramer, B.A.; Haines, J.R.; Kirchner, J.; Engholm, B.A.; Seki, M.

    1983-01-01

    Major design goals for FED-R are the achievement of: (1) a high level of neutron exposure of the test modules and (2) a capability for rapid changeout of test modules. A major factor in rapid changeout is perceived to be the location of the vacuum boundary. In FED-R this boundary was set at the first wall so that module changeout did not require the plasma chamber to be brought up to atmosphere. Efforts to realize these goals in the design resulted in a neutronically thin outboard wall for the vacuum vessel constructed of 316 stainless steel (SS) with helium as a coolant. A normalized 14-MeV neutron transmission of 0.82 is expected, with an inlet pressure of 2 MPa and a pumping power requirement of 8.7 MW. Other options considered in the study were aluminum as a wall material and water and sodium potassium (NaK) as coolants

  19. Qualitative Reliability Issues for Solid and Liquid Wall Fusion Design

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, Lee Charles

    2001-01-01

    This report is an initial effort to identify issues affecting reliability and availability of solid and liquid wall designs for magnetic fusion power plant designs. A qualitative approach has been used to identify the possible failure modes of major system components and their effects on the systems. A general set of design attributes known to affect the service reliability has been examined for the overview solid and liquid wall designs, and some specific features of good first wall design have been discussed and applied to these designs as well. The two generalized designs compare well in regard to these design attributes. The strengths and weaknesses of each design approach are seen in the comparison of specific features.

  20. Qualitative Reliability Issues for Solid and Liquid Wall Fusion Designs

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, L.C.

    2001-01-31

    This report is an initial effort to identify issues affecting reliability and availability of solid and liquid wall designs for magnetic fusion power plant designs. A qualitative approach has been used to identify the possible failure modes of major system components and their effects on the systems. A general set of design attributes known to affect the service reliability has been examined for the overview solid and liquid wall designs, and some specific features of good first wall design have been discussed and applied to these designs as well. The two generalized designs compare well in regard to these design attributes. The strengths and weaknesses of each design approach are seen in the comparison of specific features.

  1. Qualitative Reliability Issues for Solid and Liquid Wall Fusion Designs

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    2001-01-01

    This report is an initial effort to identify issues affecting reliability and availability of solid and liquid wall designs for magnetic fusion power plant designs. A qualitative approach has been used to identify the possible failure modes of major system components and their effects on the systems. A general set of design attributes known to affect the service reliability has been examined for the overview solid and liquid wall designs, and some specific features of good first wall design have been discussed and applied to these designs as well. The two generalized designs compare well in regard to these design attributes. The strengths and weaknesses of each design approach are seen in the comparison of specific features

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  3. The first IEC fusion industrial neutron generator and developments

    Science.gov (United States)

    Sved, John

    1999-06-01

    Inertial Electrostatic Confinement fusion grade plasma containment has been sporadically researched since the early 1960's. In the 1990's the work of G. H. Miley and his team at the University of Illinios, Fusion Studies Laboratory, Champaign-Urbana has stimulated a collaboration with industry. The development and test program for the first industrial IEC neutron generator has progressed to the point where an endurance test is under way to demonstrate at least 10,000 hours of operational life of the sealed chamber device without servicing. The market entry goals of steady 107 D-D n/s CW output with an air-cooled system have been achieved. DASA has invested in the development of the industrial product and the continuing basic research at the UI-FSL. The complete DASA FusionStar IEC-PS1 point source neutron generator set is described with emphasis on the interfaces to user NAA systems. The next product developments are pulsed neutron operations and higher fusion reaction rates of up to 1010 by means of affordable add-ons to the basic IEC-PS system. The production engineering experience gained will next be applied to a more challenging line source variant of the IEC. Beyond neutron and proton sources, several other IEC applications are being developed.

  4. Neutron penumbral imaging of laser-fusion targets

    International Nuclear Information System (INIS)

    Lerche, R.A.; Ress, D.B.

    1988-01-01

    Using a new technique, penumbral coded-aperture imaging, the first neutron images of laser-driven, inertial-confinement fusion targets were obtained. With these images the deuterium-tritium burn region within a compressed target can be measured directly. 4 references, 11 figures

  5. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

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

  6. Neutron excess generation by fusion neutron source for self-consistency of nuclear energy system

    International Nuclear Information System (INIS)

    Saito, Masaki; Artisyuk, V.; Chmelev, A.

    1999-01-01

    The present day fission energy technology faces with the problem of transmutation of dangerous radionuclides that requires neutron excess generation. Nuclear energy system based on fission reactors needs fuel breeding and, therefore, suffers from lack of neutron excess to apply large-scale transmutation option including elimination of fission products. Fusion neutron source (FNS) was proposed to improve neutron balance in the nuclear energy system. Energy associated with the performance of FNS should be small enough to keep the position of neutron excess generator, thus, leaving the role of dominant energy producers to fission reactors. The present paper deals with development of general methodology to estimate the effect of neutron excess generation by FNS on the performance of nuclear energy system as a whole. Multiplication of fusion neutrons in both non-fissionable and fissionable multipliers was considered. Based on the present methodology it was concluded that neutron self-consistency with respect to fuel breeding and transmutation of fission products can be attained with small fraction of energy associated with innovated fusion facilities. (author)

  7. Design of a tokamak fusion reactor first wall armor against neutral beam impingement

    International Nuclear Information System (INIS)

    Myers, R.A.

    1977-12-01

    The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem

  8. Neutron dosimetry using proportional counters with tissue equivalent walls

    International Nuclear Information System (INIS)

    Kerviller, H. de

    1965-01-01

    The author reminds the calculation method of the neutron absorbed dose in a material and deduce of it the conditions what this material have to fill to be equivalent to biological tissues. Various proportional counters are mode with walls in new tissue equivalent material and filled with various gases. The multiplication factor and neutron energy response of these counters are investigated and compared with those obtained with ethylene lined polyethylene counters. The conditions of working of such proportional counters for neutron dosimetry in energy range 10 -2 to 15 MeV are specified. (author) [fr

  9. Neutronics issues and inertial fusion energy: a summary of findings

    International Nuclear Information System (INIS)

    Latkowski, J.F.

    1998-01-01

    We have analyzed and compared five major inertial fusion energy (IFE) and two representative magnetic fusion energy (MFE) power plant designs for their environment, safety, and health (ES ampersand H) characteristics. Our work has focussed upon the neutronics of each of the designs and the resulting radiological hazard indices. The calculation of a consistent set of hazard indices allows comparisons to be made between the designs. Such comparisons enable identification of trends in fusion ES ampersand H characteristics and may be used to increase the likelihood of fusion achieving its full potential with respect to ES ampersand H characteristics. The present work summarizes our findings and conclusions. This work emphasizes the need for more research in low-activation materials and for the experimental measurement of radionuclide release fractions under accident conditions

  10. Neutron dosimetry for radiation damage in fission and fusion reactors

    International Nuclear Information System (INIS)

    Smith, D.L.

    1979-01-01

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

  11. Role of neutrons in the fusion of nuclei

    International Nuclear Information System (INIS)

    Zagrebaev, V. I.; Samarin, V. V.

    2007-01-01

    The problem of a quantum-mechanical description of a near-barrier fusion of heavy nuclei under conditions of a strong coupling of their relative motion to neutron degrees of freedom is studied. New mutually complementing efficient models employing the multidimensional time-dependent Schroedinger equation and methods for numerically solving the problems in question within these models are proposed. This makes it possible to study the evolution of wave functions for outer neutrons and to calculate nucleon transfer probabilities and occupation numbers for quasimolecular states over the entire energy range, including that of deep-subbarrier energies. It is shown that the valence-neutron wave functions extend over the volumes of both colliding nuclei before their surfaces come into contact and even before these nuclei overcome the Coulomb barrier (collectivization of outer nucleons). This gives rise to a substantial neutron-transfer effect on the fusion of nuclei and, in particular, to a sizable increase in the probability of subbarrier fusion for specific combinations of nuclei

  12. A comparison of microstructures in copper irradiated with fission, fusion, and spallation neutrons

    International Nuclear Information System (INIS)

    Muroga, T.; Heinisch, H.L.; Sommer, W.F.; Ferguson, P.D.

    1992-01-01

    The objective of this work is to investigate the effects of the neutron energy spectrum in low dose irradiations on the microstructure and mechanical properties of metals. The microstructures of pure copper irradiated to low doses at 36-90 C with spallation neutrons, fusion neutrons and fission neutrons are compared. The defect cluster densities for the spallation and fusion neutrons are very similar when compared on the basis of displacements per atom (dpa). In both cases, the density increases in proportion to the square root of the dpa. The difference in defect density between fusion neutrons and fission neutrons corresponds with differences observed in data on yield stress changes

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  14. Behaviour of neutrons passing through the Bloch wall

    International Nuclear Information System (INIS)

    Schaerpf, O.

    1976-01-01

    In part I of the present paper the pertinent knowledge about Bloch walls is presented and developed insofar as it appears necessary for the experiments with neutrons, that is to say the direction of magnetization within the domains, the calculation of the variation of magnetization in the wall, the wall thickness, and the zigzag structure of the Bloch wall. In part II it is first clarified why the Bloch wall can be treated as a continuum problem. It shows that this is possible far away from Laue reflexes. For angles far away from Laure-reflex angles the interaction of the periodic structure of the magnetization can be described with the aid of an averaged magnetic flux density. The consequence of it is the possibility of treating the problem by means of a Schroedinger equation with continous interaction. This leads to a law of refraction. The question of the possibilities for explaining the intensity behavior is treated in part III. This part, from different aspects, describes the fact, which already was pointed out in Schaerpf, O., Vehoff, H., Schwink, Ch. 1973, that the spin of the neutrons in passing through the wall is partly taken along by the magnetization gradually rotating in the wall. (orig./WBU) [de

  15. Novel neutralized-beam intense neutron source for fusion technology development

    International Nuclear Information System (INIS)

    Osher, J.E.; Perkins, L.J.

    1983-01-01

    We describe a neutralized-beam intense neutron source (NBINS) as a relevant application of fusion technology for the type of high-current ion sources and neutral beamlines now being developed for heating and fueling of magnetic-fusion-energy confinement systems. This near-term application would support parallel development of highly reliable steady-state higher-voltage neutral D 0 and T 0 beams and provide a relatively inexpensive source of fusion neutrons for materials testing at up to reactor-like wall conditions. Beam-target examples described incude a 50-A mixed D-T total (ions plus neutrals) space-charge-neutralized beam at 120 keV incident on a liquid Li drive-in target, or a 50-A T 0 + T + space-charge-neutralized beam incident on either a LiD or gas D 2 target with calculated 14-MeV neutron yields of 2 x 10 15 /s, 7 x 10 15 /s, or 1.6 x 10 16 /s, respectively. The severe local heat loading on the target surface is expected to limit the allowed beam focus and minimum target size to greater than or equal to 25 cm 2

  16. Fusion-neutron effects on magnetoresistivity of copper stabilizer materials

    International Nuclear Information System (INIS)

    Guinan, M.W.; Van Konynenburg, R.A.

    1983-01-01

    The objective of this work is to quantify the changes which occur in the magnetoresistivity of coppers (having various purities and pretreatments, and at magnetic fields up to 12 T during the course of sequential fusion neutron irradiations at about 4 0 K and anneals to room temperature. In conjunction with work in progress by Coltman and Klabunde of ORNL, the results should lead to engineering design data for the stabilizers of superconducting magnets in fusion reactors. These magnets are expected to be irradiated during reactor operation and warmed to room temperature periodically during maintenance

  17. Fusion-neutron effects on magnetoresistivity of copper stabilizer materials

    Energy Technology Data Exchange (ETDEWEB)

    Guinan, M.W.; Van Konynenburg, R.A.

    1983-02-24

    The objective of this work is to quantify the changes which occur in the magnetoresistivity of coppers (having various purities and pretreatments, and at magnetic fields up to 12 T during the course of sequential fusion neutron irradiations at about 4/sup 0/K and anneals to room temperature. In conjunction with work in progress by Coltman and Klabunde of ORNL, the results should lead to engineering design data for the stabilizers of superconducting magnets in fusion reactors. These magnets are expected to be irradiated during reactor operation and warmed to room temperature periodically during maintenance.

  18. Transport of D-D fusion neutrons in thick concrete

    International Nuclear Information System (INIS)

    Ku, L.P.; Kolibal, J.G.

    1982-07-01

    By altering the collision mechanism in the numerical transport calculations, and by constructing an analytical model based on age-diffusion theory, the outstanding feature in the life history of D-D fusion neutrons penetrating deeply into ordinary concrete is shown to be the transport in the 2.3 MeV oxygen anti-resonance. This result is used to assess the impact of the cross-section uncertainties and the uncertainties due to variations in the D-D fusion spectrum and temperature

  19. Beryllium neutron activation detector for pulsed DD fusion sources

    International Nuclear Information System (INIS)

    Talebitaher, A.; Springham, S.V.; Rawat, R.S.; Lee, P.

    2011-01-01

    A compact fast neutron detector based on beryllium activation has been developed to perform accurate neutron fluence measurements on pulsed DD fusion sources. It is especially well suited to moderate repetition-rate ( 9 Be(n,α) 6 He cross-section, energy calibration of the proportional counters, and numerical simulations of neutron interactions and beta-particle paths using MCNP5. The response function R(E n ) is determined over the neutron energy range 2-4 MeV. The count rate capability of the detector has been studied and the corrections required for high neutron fluence measurements are discussed. For pulsed DD neutron fluencies >3×10 4 cm -2 , the statistical uncertainty in the fluence measurement is better than 1%. A small plasma focus device has been employed as a pulsed neutron source to test two of these new detectors, and their responses are found to be practically identical. Also the level of interfering activation is found to be sufficiently low as to be negligible.

  20. Neutronics design for a spherical tokamak fusion-transmutation reactor

    International Nuclear Information System (INIS)

    Deng Meigen; Feng Kaiming; Yang Bangchao

    2002-01-01

    Based on studies of the spherical tokamak fusion reactors, a concept of fusion-transmutation reactor is put forward. By using the one-dimension transport and burn-up code BISON3.0 to process optimized design, a set of plasma parameters and blanket configuration suitable for the transmutation of MA (Minor Actinides) nuclear waste is selected. Based on the one-dimension calculation, two-dimension calculation has been carried out by using two-dimension neutronics code TWODANT. Combined with the neutron flux given by TWODANT calculation, burn-up calculation has been processed by using the one-dimension radioactivity calculation code FDKR and some useful and reasonable results are obtained

  1. 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

  2. Design of neutron streak camera for fusion diagnostics

    International Nuclear Information System (INIS)

    Wang, C.L.; Kalibjian, R.; Singh, M.S.

    1982-06-01

    The D-T burn time for advanced laser-fusion targets is calculated to be very short, 2 . Each fission fragment leaving the cathode generates 400 secondary electrons that are all < 20 eV. These electrons are focussed to a point with an extractor and an anode, and are then purified with an electrostatic deflector. The electron beam is streaked and detected with the standard streak camera techniques. Careful shielding is needed for x-rays from the fusion target and general background. It appears that the neutron streak camera can be a viable and unique tool for studying temporal history of fusion burns in D-T plasmas of a few keV ion temperature

  3. The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

    International Nuclear Information System (INIS)

    Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

    2012-01-01

    In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

  4. Computational modeling of the axial-cylindrical inertial electrostatic confinement fusion neutron generator

    Science.gov (United States)

    Bromley, Blair Patrick

    2001-12-01

    The axial-cylindrical Inertial Electrostatic Confinement fusion neutron generator (IEC C-Device) is a high- voltage, low-pressure glow discharge device that produces neutrons from the deuterium-deuterium fusion reaction. Such a neutron source has potential applications for neutron activation analysis and capture therapies for cancer treatment. The IEC C-Device operating with deuterium fuel is modeled with the CHIMP computer code developed and written completely by the author to predict the fusion neutron generation rate and the plasma physics behavior using fundamental first principles. The CHIMP code is a time-dependent, spatially two-dimensional (r,z), particle-in-cell, Monte-Carlo-Collision (PIC-MCC) direct simulation model. The effects of secondary electron emission due to ion and electron impact on the metal electrodes and the glass walls and charge build-up on the glass wall are included. Either monatomic or molecular ions and electrons are modeled in a monatomic or molecular background neutral deuterium gas. CHIMP code predictions are compared against experimental results for the C-Device operating between 10 and 30 kV of anode voltage, between 10 and 40 mA of electrode current, and between 0.29 and 1.1 milliTorr of deuterium gas pressure. A calibration factor for the pressure accounts for the calibration of the ionization pressure gauge in the experiment, and an estimated pressure drop between the main chamber of the C-Device and the pressure gauge that is downstream of the exhaust port. Upgraded versions of the CHIMP code which have modifications to the algorithms for the boundary conditions, and which include charge exchange processes, and the contribution of fast neutrals to the neutron generation rate are also tested against several experimental data points. Although the CHIMP code gives predictions for the neutron generation rate that exhibit the same near-linear trends with current found in the experiment, it is apparent that at least five types of

  5. Neutron streaming evaluation for the DREAM fusion power reactor

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  6. Fusion materials high energy-neutron studies. A status report

    International Nuclear Information System (INIS)

    Doran, D.G.; Guinan, M.W.

    1980-01-01

    The objectives of this paper are (1) to provide background information on the US Magnetic Fusion Reactor Materials Program, (2) to provide a framework for evaluating nuclear data needs associated with high energy neutron irradiations, and (3) to show the current status of relevant high energy neutron studies. Since the last symposium, the greatest strides in cross section development have been taken in those areas providing FMIT design data, e.g., source description, shielding, and activation. In addition, many dosimetry cross sections have been tentatively extrapolated to 40 MeV and integral testing begun. Extensive total helium measurements have been made in a variety of neutron spectra. Additional calculations are needed to assist in determining energy dependent cross sections

  7. Fiscal year 1976 DT fusion neutron irradiations and dosimetry at the LLL rotating target neutron source

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation of 319 samples during 19 irradiation periods (beam-on time of more than 1026 hours) is described. Experiments from 24 individuals representing 11 institutions are summarized. The numbers of the UCID dosimetry reports detailing each of the irradiations are given

  8. Thermostructural design of the first wall/blanket for the TITAN-RFP fusion reactor

    International Nuclear Information System (INIS)

    Orient, G.E.; Blanchard, J.P.; Ghoniem, N.M.

    1987-01-01

    The mass power density, which is defined as the average power per unit mass within the magnet boundary, is a rough and general measure of economic competitiveness. Conn et al. (1985) have identified a target value of 100 kW(e)/tonne as a reasonable threshold for 'compact' commercial fusion systems. In pursuit of this goal, Hagenson et al. (1984) and Najmabadi et al. (1987) have pointed out the inherent characteristics of the RFP toroidal confinement concept which allow it to exceed this target value. It is inevitable that the compactness of the fusion power core will introduce a unique set of design issues. The special design concerns stem from high thermal surface fluxes, high bulk energy deposition by neutrons, and a relatively short blanket structural lifetime. In the TITAN-RFP, study Najmabadi et al. (1987) investigate a number of blanket (B) and first wall (FW) options suitable for high power density fusion reactors. Final choices were made for two designs: A high pressure aqueous blanket and a vanadium/lithium self-cooled blanket. The first design utilizes a pressurized aqueous loop containing a lithium compound dissolved in water, while the second design is based upon a self-cooled lithium-vanadium blanket. In this paper, we consider the beginning-of-life (BOL) thermostructural design and analysis of only the second concept. (orig./GL)

  9. Deuterium behavior in first-wall materials for nuclear fusion

    International Nuclear Information System (INIS)

    Bernard, E.

    2012-01-01

    Plasma-wall interactions play an important part while choosing materials for the first wall in future fusion reactors. Moreover, the use of tritium as a fuel will impose safety limits regarding the total amount present in the tokamak. Previous analyses of first-wall samples exposed to fusion plasma highlighted an in-bulk migration of deuterium (as an analog to tritium) in carbon materials. Despite its limited value, this retention is problematic: contrary to co-deposited layers, it seems very unlikely to recover easily the deuterium retained in such a way. Because of the difficult access to in situ samples, most published studies on the subject were carried out using post-mortem sample analysis. In order to access to the dynamic of the phenomenon and come apart potential element redistribution during storage, we set up a bench intended for simultaneous low-energy ion implantation, reproducing the deuterium interaction with first-wall materials, and high-energy micro beam analysis. Nuclear reaction analysis performed at the micrometric scale (μNRA) allows to characterize deuterium repartition profiles in situ. This analysis technique was confirmed to be non-perturbative of the mechanisms studied. We observed on the experimental data set that the material surface (0-1 μm) display a high and nearly constant deuterium content, with a uniform distribution. On the contrary, in-bulk deuterium (1-11 μm) localizes in preferential trapping sites related to the material microstructure. In-bulk deuterium inventory seems to increase with the incident fluence, in spite of the wide data scattering attributed to the structure variation of studied areas. Deuterium saturation at the surface as well as in-depth migration are instantaneous; in-vacuum storage leads to a small deuterium global desorption. Observations made via μNRA were coupled with results from other characterization techniques. X-ray μtomography allowed to identify porosities as the preferential trapping sites

  10. Nuclear fusion ion beam source composed of optimum channel wall

    International Nuclear Information System (INIS)

    Furukaw, T.

    2007-01-01

    Full text of publication follows: Numerical and experimental researches of the hall-type beam accelerator was conducted by highlighting both neutral species and material of acceleration channel wall. The hall-type beam accelerator is expected as ion beam source for nuclear fusion since it could product ion beam density over 10 3 times as high as that of electrostatic accelerator, which is used regularly as beam heating device, because it is proven that the beam heating method could accelerate ion to high energy beam by electric field and heat plasma to ultra high temperature of 100 million degrees or more. At high-voltage mode of DC regime that is normal operational condition, however, the various plasma MHD (magneto-hydrodynamic) instabilities are generated. In particular, the large-amplitude and low-frequency plasma MHD instability in the tens of kHz among them has been a serious problem that should be solved to improve the operational stability and the system durability. So, we propose a hall-type beam accelerator with new design concepts; both acquisition of simultaneous solution for reducing the plasma MHD instability and the accelerator core overheating and optimum combination of the acceleration channel wall material. The technologies for this concept are as follows: 1) To increase neutral species velocity-inlet in acceleration channel by preheating propellant through circularly propellant conduit line inside accelerator system could bring about the lower amplitude of the instability. 2) Through this method, the accelerator system is cooled, and the higher thrust and specific-impulse is produced with hardly changing thrust efficiency at the same time. 3) To select BN (Boron- Nitride) and Al 2 O 3 as wall material of ionization- and acceleration-zone in acceleration channel respectively having different secondary-electron emission-coefficient could achieve the higher-efficiency and -durability. The hall-type beam accelerator designed using these technologies

  11. Erosion and redeposition at the vessel walls in fusion devices

    International Nuclear Information System (INIS)

    Naujoks, D.; Behrisch, R.

    1995-01-01

    The plasma induced erosion and redeposition at the vessel walls in today's fusion devices have been investigated both with the computer simulation code ERO, and in experiments. Well prepared carbon probes with implanted and evaporated markers in the surface layers have been exposed in the scrape-off layer (SOL) of several tokamaks such as JET, TEXTOR and ASDEX-Upgrade. The main plasma parameters (electron density and temperature, impurity concentration in the SOL) are simultaneously determined. After exposure to single plasma discharges, erosion and redeposition of the marker material were measured by surface layer analysis with MeV ion beam techniques. The experimental results were compared with the results from the ERO code. The measured erosion/redeposition could be described with ERO, which takes into account the impurity concentration in the SOL, the dynamical change of the surface composition (causing a modification of the sputtering yield during the exposure) and ExB drift effects. ((orig.))

  12. Towards a strategy of reliable fusion first-wall design

    International Nuclear Information System (INIS)

    Schultz, J.H.

    1981-05-01

    Fusion first walls are subject to a large number of possible failure mechanisms, including erosion due to sputtering, arcing, blistering and vaporization and crack growth due to thermal and magnetic stresses. Each of these failure mechanisms is poorly characterized and has the potential of being severe. A strategy for designing reliably in the face of great uncertainty is discussed. Topological features beneficial to reactor availability are identified. The integration of limiter pumping with rf wave launching is discussed, as a means of simplifying reactor design. The concept of a sewer limiter is introduced, as a possible long-life limiter topology. The concept of flexible armor is discussed, as a means of extending maximum life

  13. OVERVIEW OF NEUTRON MEASUREMENTS IN JET FUSION DEVICE.

    Science.gov (United States)

    Batistoni, P; Villari, R; Obryk, B; Packer, L W; Stamatelatos, I E; Popovichev, S; Colangeli, A; Colling, B; Fonnesu, N; Loreti, S; Klix, A; Klosowski, M; Malik, K; Naish, J; Pillon, M; Vasilopoulou, T; De Felice, P; Pimpinella, M; Quintieri, L

    2017-10-05

    The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  14. DT fusion neutron irradiation of ORNL magnesium oxide crystals and BNL--LASL superconductor wires

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1978-01-01

    The DT fusion neutron irradiation of two ORNL magnesium oxide crystals and eleven BNL-LASL superconductor wires is described. The sample position and neutron dose record are given. The maximum neutron fluence on any sample was 2.16 x 10 16 neutrons/cm 2

  15. Irradiation capsule for testing magnetic fusion reactor first-wall materials at 60 and 2000C

    International Nuclear Information System (INIS)

    Conlin, J.A.

    1985-08-01

    A new type of irradiation capsule has been designed, and a prototype has been tested in the Oak Ridge Research Reactor (ORR) for low-temperature irradiation of Magnetic Fusion Reactor first-wall materials. The capsule meets the requirements of the joint US/Japanese collaborative fusion reactor materials irradiation program for the irradiation of first-wall fusion reactor materials at 60 and 200 0 C. The design description and results of the prototype capsule performance are presented

  16. Dynamic loading of the structural wall in a lithium fall fusion reactor

    International Nuclear Information System (INIS)

    Glenn, L.A.

    1979-01-01

    In one version of an inertial confinement fusion (ICF) power reactor, the laser-imploded pellet is surrounded by a thick, annular 'waterfall' of liquid lithium. The fall has three functions: to breed tritium for pellet resupply, to act as an energy sink and heat exchange mdeium with an external power loop, and to protect the first wall of the reactor from excessive neutronic and hydrodynamic loading. Our primary concern here is with this last function. We formulated a simple model of a lithium-fall ICF reactor and calculated the fall disassembly and the subsequent fluid-wall interaction resulting from the energy deposition by the imploded pellet. Two potential mechanisms for wall damage were identified: surface erosion and hoop failure. For single fall designs, the erosion problem appears to be serious. Concentric annuli (multiple fall) or packed jet configurations may be feasible but experiments are needed to clarify the physical model, especially with reg (orig.)ard to /orig.the characteristics of the cavitated liquid lithium and of the two-phase liquid-vapor region.

  17. Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

    International Nuclear Information System (INIS)

    Ku, L.P.; Price, W.G. Jr.

    1977-08-01

    The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model

  18. Novel, spherically-convergent ion systems for neutron source and fusion energy production

    International Nuclear Information System (INIS)

    Barnes, D.C.; Nebel, R.A.; Ribe, F.L.; Schauer, M.M.; Schranck, L.S.; Umstadter, K.R.

    1999-01-01

    Combining spherical convergence with electrostatic or electro-magnetostatic confinement of a nonneutral plasma offers the possibility of high fusion gain in a centimeter-sized system. The physics principles, scaling laws, and experimental embodiments of this approach are presented. Steps to development of this approach from its present proof-of-principle experiments to a useful fusion power reactor are outlined. This development path is much less expensive and simpler, compared to that for conventional magnetic confinement and leads to different and useful products at each stage. Reactor projections show both high mass power density and low to moderate wall loading. This approach is being tested experimentally in PFX-I (Penning Fusion eXperiment-Ions), which is based on the following recent advances: 1) Demonstration, in PFX (our former experiment), that it is possible to combine nonneutral electron plasma confinement with nonthermal, spherical focussing; 2) Theoretical development of the POPS (Periodically Oscillating Plasma Sphere) concept, which allows spherical compression of thermal-equilibrium ions; 3) The concept of a massively-modular approach to fusion power, and associated elimination of the critical problem of extremely high first wall loading. PFX-I is described. PFX-I is being designed as a small (<1.5 cm) spherical system into which moderate-energy electrons (up to 100 kV) are injected. These electrons are magnetically insulated from passing to the sphere and their space charge field is then used to spherically focus ions. Results of initial operation with electrons only are presented. Deuterium operation can produce significant neutron output with unprecedented efficiency (fusion gain Q). copyright 1999 American Institute of Physics

  19. State-of-the-art 3-D neutronics analysis methods for fusion energy systems

    International Nuclear Information System (INIS)

    Wilson, P.P.H.; Feder, R.; Fischer, U.; Loughlin, M.; Petrizzi, L.; Wu, Y.

    2007-01-01

    Recent advances in radiation transport simulation tools enable an increased fidelity and accuracy in modeling complex geometries in fusion systems. Future neutronics calculations for design and analysis will increasingly be based directly on 3-D CAD-based geometries, allowing enhanced model complexity, reduced human effort and improved quality assurance. Improvements have been made in both stochastic and deterministic radiation transport methodologies. To adapt the MCNP stochastic transport software, the translator approach allows CAD geometries to be converted from their native formats into standard input files, while the direct geometry approach uses computer graphics algorithms to perform the radiation transport on the CAD geometry itself. The former takes advantage of the efficiency of the native MCNP software without modifications while the latter permits the modeling of more complex surfaces. The ATTILA radiation transport package uses a finite-element formulation of the discrete-ordinate methodology to provide a deterministic solution on a tetrahedral mesh derived automatically from a CAD-based geometry. All of these tools are being applied to a dedicated benchmark problem consisting of a 40 degree sector of the ITER machine defined only in a CAD-based solid model. The specific benchmark problems exercise the ability to use a CAD-based geometry to solve a range of fusion neutronics problems including neutron wall loading, deep penetration and narrow duct streaming. The results of this exercise will be used to validate/qualify these tools for use on ITER. At the same time, many of these tools are being used to support the design of ITER components and other related fusion systems. UW has provided high-fidelity nuclear analysis of ITER first wall and shield modules identifying local effects of geometric features. ASIPP has used the MCAM tool to update and extend the existing ITER basic model and used it for neutronics analysis of the proposed Chinese ITER

  20. State-of-the-art 3-D neutronics analysis methods for fusion energy systems

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, P.P.H. [Wisconsin-Madison Univ., Madison, WI (United States); Feder, R. [Princeton Plasma Physics Lab. (United States); Fischer, U. [Forschungszentrum Karlsruhe (Germany); Loughlin, M. [United Kingdom Atomic Energy Authority (United Kingdom); Petrizzi, L. [ENEA-Frascati (Italy); Wu, Y. [Academy of Sciences (China). Inst. of Plasma Physics; Youssef, M. [California Univ., Los Angeles, CA (United States)

    2007-07-01

    Recent advances in radiation transport simulation tools enable an increased fidelity and accuracy in modeling complex geometries in fusion systems. Future neutronics calculations for design and analysis will increasingly be based directly on 3-D CAD-based geometries, allowing enhanced model complexity, reduced human effort and improved quality assurance. Improvements have been made in both stochastic and deterministic radiation transport methodologies. To adapt the MCNP stochastic transport software, the translator approach allows CAD geometries to be converted from their native formats into standard input files, while the direct geometry approach uses computer graphics algorithms to perform the radiation transport on the CAD geometry itself. The former takes advantage of the efficiency of the native MCNP software without modifications while the latter permits the modeling of more complex surfaces. The ATTILA radiation transport package uses a finite-element formulation of the discrete-ordinate methodology to provide a deterministic solution on a tetrahedral mesh derived automatically from a CAD-based geometry. All of these tools are being applied to a dedicated benchmark problem consisting of a 40 degree sector of the ITER machine defined only in a CAD-based solid model. The specific benchmark problems exercise the ability to use a CAD-based geometry to solve a range of fusion neutronics problems including neutron wall loading, deep penetration and narrow duct streaming. The results of this exercise will be used to validate/qualify these tools for use on ITER. At the same time, many of these tools are being used to support the design of ITER components and other related fusion systems. UW has provided high-fidelity nuclear analysis of ITER first wall and shield modules identifying local effects of geometric features. ASIPP has used the MCAM tool to update and extend the existing ITER basic model and used it for neutronics analysis of the proposed Chinese ITER

  1. Ion beam pellet fusion as a CTR neutron test source

    International Nuclear Information System (INIS)

    Arnold, R.; Martin, R.

    1975-07-01

    Pellet fusion, driven by nanosecond pulses containing α particles with 200 MeV energy, is being developed as a neutron source. A prototype system is in the conceptual design stage. During the coming year, engineering design of required accelerator components, storage rings, and pellet configurations, as well as experiments on energy deposition mechanisms, should be accomplished. Successful construction and tests of prototype rings, followed by two years of full scale system construction, would give a source producing a useful flux of fusion neutrons for materials testing. The system as currently envisioned would employ 100 small superconducting high field storage rings (15 cm radius, 140 kG field) which would be synchronously filled with circulating 1 nsec pulses from a 200 MeV linear accelerator over a period of 3 x 10 -4 sec. These ion pulses would all be simultaneously extracted, forming a total current of 10 kA, and focussed from all directions on a deuterium and tritium (DT) pellet with 0.17 mm radium, surrounded by a heavier (metal) coating to increase confinement time and aid compression efficiency. The overall repetition rate, limited principally by physical transport of the pellets, could reach 100/sec. Spacing between pellet and focussing elements would be about 1 m. The predominant engineering problems are the fast extraction mechanism and beam transport devices for the storage rings. Additional theoretical and experimental studies are required on the crucial energy deposition and transport mechanisms in pellets with ion beam heating before firm estimates can be given. Preliminary estimates suggest fusion neutron yields of at least 10 14 /sec and possibly 10 16 /sec are possible, with optimal pellet dynamics, but without the necessity for any large advances in the state-of-the-art in accelerator and storage ring design. (auth)

  2. Characterization of graded iron / tungsten layers for the first wall of fusion reactors

    International Nuclear Information System (INIS)

    Heuer, Simon

    2017-01-01

    The nuclear fusion has great potential to enable a CO 2 -neutral energy supply of future generations. The technical utilization of this energy source has hitherto been a challenge. In particular, high thermal loads and neutron-induced damage lead to extreme demands on the choice of materials for plasma-facing components (PFCs). These are therefore, as currently understood, made from a tungsten protective layer which is joined to a structure of low activation ferritic-martensitic (LAFM) steel. Due to the discrete transition of material properties at the LAFM-W joining zone as well as thermal loads, macroscopic stresses and plastic strains arise here. A feasible way to reduce this is to implement an intermediate layer with graded LAFM / W ratio, a so-called functional graded material (FGM). In the present work, macro-stresses and strains in the first wall of the fusion reactor DEMO are examined and evaluated by means of a finite element simulation. In this framework model components with and without graded interlayer are taken into account and the advantage of a FGM is emphasized. Parameter studies serve as a constructive guideline for the structural implementation of FGMs and components of the first wall. In addition, the feasibility of four methods (magnetron sputtering, liquid phase infiltration, modified atmospheric plasma spraying and electrodischarge sintering) with respect to the fabrication of FGMs is being studied. The resulting layers are microstructurally, thermo-physically and mechanically examined in detail. Based on this characterization and the finite element simulation, their suitability as a graded layer in the first wall of DEMO is evaluated and finally compared with alternative joining systems that are currently being tested in the research environment. [de

  3. First fusion neutrons from a thermonuclear weapon device

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    An account of the first observation of thermonuclear neutrons from a hydrogen weapon, the George shot, is presented. A personal narrative by the researchers J. Allred and L. Rosen includes such topics as the formation of the experimental team, description of the experimental technique, testing the experimental apparatus, testing the effects of a blast, a description of the test area, and the observation of neutrons from fusion. Excerpts are presented from several chapters of the Scientific Director's report on the atomic weapons tests of 1951. Also included is a brief description of the basic design of the hydrogen bomb, a recounting of subsequent developments, and short scientific biographies of the researchers. 21 figures, 2 tables

  4. Solid state detectors for neutron radiation monitoring in fusion facilities

    International Nuclear Information System (INIS)

    Gómez-Ros, J.M.

    2014-01-01

    The purpose of this communication is to summarize the main solid state based detectors proposed for neutron diagnostic in fusion applications and their applicability under the required harsh conditions in terms of intense radiation, high temperature and available space restrictions. Activation systems, semiconductor based detectors, luminescent materials and Cerenkov fibre optics sensors (C-FOS) are the main devices that are described. - Highlights: • A state-of-the-art summary of solid state based detectors are described. • Conditions and restrictions for their applicability are described. • A list of the 38 more relevant references has been included

  5. Materials for heat flux components of the first wall in fusion reactors

    International Nuclear Information System (INIS)

    Hoven, H.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

    1985-08-01

    Materials of the First Wall in near-fusion plasma machines are subjected to a complex load system resulting from the plasma-wall interaction. The materials for their part also influence the plasma. Suitable materials must be available in order to ensure that the wall components achieve a sufficiently long dwell time and that their effects on the plasma remain small and controllable. The present report discusses relations between the plasma-wall interaction, the reactions of the materials and testing and examination methods for specific problems in developing and selecting suitable materials for highly stressed components on the First Wall of fusion reactors. (orig.)

  6. Measurements of fusion neutron yields by neutron activation technique: Uncertainty due to the uncertainty on activation cross-sections

    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); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Via E. Fermi, 45, 00044 Frascati, Rome (Italy); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Sjöstrand, Henrik; Conroy, Sean [Department of Physics and Astronomy, Uppsala University, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2015-07-11

    The neutron activation technique is routinely used in fusion experiments to measure the neutron yields. This paper investigates the uncertainty on these measurements as due to the uncertainties on dosimetry and activation reactions. For this purpose, activation cross-sections were taken from the International Reactor Dosimetry and Fusion File (IRDFF-v1.05) in 640 groups ENDF-6 format for several reactions of interest for both 2.5 and 14 MeV neutrons. Activation coefficients (reaction rates) have been calculated using the neutron flux spectra at JET vacuum vessel, both for DD and DT plasmas, calculated by MCNP in the required 640-energy group format. The related uncertainties for the JET neutron spectra are evaluated as well using the covariance data available in the library. These uncertainties are in general small, but not negligible when high accuracy is required in the determination of the fusion neutron yields.

  7. An investigation on detection and measurement of fusion neutron spectrum and radiation flux in large tokamak

    International Nuclear Information System (INIS)

    Yang Jinwei; Li Wenzhong; Zhang Wei

    2003-01-01

    The detection methods, detectors and spectrometers of D-D and D-T fusion neutron have been overviewed in large tokamaks. Some options are proposed for developing new detection systems of fusion neutrons suitable to the HL-2A tokamak. (authors)

  8. Fusion neutron generation by high-repetitive target injection

    International Nuclear Information System (INIS)

    Kitagawa, Yoneyoshi

    2015-01-01

    Pellet injection and repetitive laser illumination are key technologies for realizing inertial fusion energy. The Graduate School for the Creation of New Photonics Industries, Hamamatsu Photonics K. K. and Toyota Motor Corporation demonstrate the pellet injection, counter laser beams' engagement and neutron generation. Deuterated polystyrene (CD) bead pellets, after free-falling for a distance of 18 cm at 1 Hz, are successfully engaged by two counter laser beams from a diode-pumped, ultra-intense laser HAMA. The laser energy, pulse duration, wavelength and the intensity are 0.63 J per beam, 104 fs, 811 nm and 4.7 x 10 18 W/cm 2 , respectively. The irradiated pellets produce D (D, n) 3 He-reacted neutrons with a maximum yield of 9.5 x 10 4 /4π sr/shot. A straight channel with 10 μm-diameter is found through the beads. The pellet size is 1 mm. The results indicate potentially useful technologies for the next step in realizing inertial fusion energy. The results are reviewed as well as some oversea activities. (author)

  9. Neutron Generation by Laser-Driven Spherically Convergent Plasma Fusion

    Science.gov (United States)

    Ren, G.; Yan, J.; Liu, J.; Lan, K.; Chen, Y. H.; Huo, W. Y.; Fan, Z.; Zhang, X.; Zheng, J.; Chen, Z.; Jiang, W.; Chen, L.; Tang, Q.; Yuan, Z.; Wang, F.; Jiang, S.; Ding, Y.; Zhang, W.; He, X. T.

    2017-04-01

    We investigate a new laser-driven spherically convergent plasma fusion scheme (SCPF) that can produce thermonuclear neutrons stably and efficiently. In the SCPF scheme, laser beams of nanosecond pulse duration and 1 014- 1 015 W /cm2 intensity uniformly irradiate the fuel layer lined inside a spherical hohlraum. The fuel layer is ablated and heated to expand inwards. Eventually, the hot fuel plasmas converge, collide, merge, and stagnate at the central region, converting most of their kinetic energy to internal energy, forming a thermonuclear fusion fireball. With the assumptions of steady ablation and adiabatic expansion, we theoretically predict the neutron yield Yn to be related to the laser energy EL, the hohlraum radius Rh, and the pulse duration τ through a scaling law of Yn∝(EL/Rh1.2τ0.2 )2.5. We have done experiments at the ShengGuangIII-prototype facility to demonstrate the principle of the SCPF scheme. Some important implications are discussed.

  10. Ion-bombardment effects on the fatigue life of stainless steel under simulated fusion first-wall conditions

    International Nuclear Information System (INIS)

    Kohse, G.E.

    1983-02-01

    An experiment which uses the MITR-II 5 MW research reactor to simulate several aspects of the anticipated environment of a fusion reactor first wall is described. Pressurized tube specimens are subjected simultaneously to stress and temperature cycling, surface bombardment by energetic helium and lithium ions and bulk irradiation by high-energy neutrons. Analysis of the samples is aimed primarily at determining the behavior of the ion bombarded surface layer, which has a depth of 2.5 μm, with particular reference to possible effects on the fatigue life of the material

  11. Fusion neutron yield and flux calculation on HT-7 superconducting tokamak

    International Nuclear Information System (INIS)

    Fu Yanzhang; Zhu Yubao; Chen Juequan

    2006-01-01

    Neutron yield and flux have been numerically estimated on HT-7 tokamak. The total fusion neutron yield and neutron flux distribution on different positions and azimuth angles of the device are presented. Analyses on the errors induced by ion temperature and density distribution factors are given in detail. The results of the calculations provide a useful database for neutron diagnostics and neutron radiation protection. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-12-01

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

  13. Tungsten as First Wall Material in Fusion Devices

    International Nuclear Information System (INIS)

    Kaufmann, M.

    2006-01-01

    In the PLT tokamak with a tungsten limiter strong cooling of the central plasma was observed. Since then mostly graphite has been used as limiter or target plate material. Only a few tokamaks (limiter: FTU, TEXTOR; divertor: Alcator C-Mod, ASDEX Upgrade) gained experience with high-Z-materials. With the observed strong co- deposition of tritium together with carbon in JET and as a result of design studies of fusion reactors, it became clear that in the long run tungsten is the favourite for the first-wall material. Tungsten as a plasma facing material requires intensive research in all areas, i.e. in plasma physics, plasma wall-interaction and material development. Tungsten as an impurity in the confined plasma reveals considerable differences to carbon. Strong radiation at high temperatures, in connection with mostly a pronounced inward drift forms a particular challenge. Turbulent transport plays a beneficial role in this regard. The inward drift is an additional problem in the pedestal region of H-mode plasmas in ITER-like configurations. The erosion by low energy hydrogen atoms is in contrast to carbon small. However, erosion by fast particles from heating measures and impurity ions, accelerated in the sheath potential, play an important role in the case of tungsten. Radiation by carbon in the plasma boundary reduces the load to the target plates. Neon or Argon as substitutes will increase the erosion of tungsten. So far experiments have demonstrated that in most scenarios the tungsten content in the central plasma can be kept sufficiently small. The material development is directed to the specific needs of existing or future devices. In ASDEX Upgrade, which will soon be a divertor experiment with a complete tungsten first-wall, graphite tiles are coated with tungsten layers. In ITER, the solid tungsten armour of the target plates has to be castellated because of its difference in thermal expansion compared to the cooling structure. In a reactor the technical

  14. Neutron and photon transport calculations in fusion system. 2

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    On the application of MCNP to the neutron and {gamma}-ray transport calculations for fusion reactor system, the wide range design calculation has been carried out in the engineering design activities for the international thermonuclear fusion experimental reactor (ITER) being developed jointly by Japan, USA, EU and Russia. As the objects of shielding calculation for fusion reactors, there are the assessment of dose equivalent rate for living body shielding and the assessment of the nuclear response for the soundness of in-core structures. In the case that the detailed analysis of complicated three-dimensional shapes is required, the assessment using MCNP has been carried out. Also when the nuclear response of peripheral equipment due to the gap streaming between blanket modules is evaluated with good accuracy, the calculation with MCNP has been carried out. The analyses of the shieldings for blanket modules and NBI port are explained, and the examples of the results of analyses are shown. In the blanket modules, there are penetrating holes and continuous gap. In the case of the NBI port, shielding plug cannot be installed. These facts necessitate the MCNP analysis with high accuracy. (K.I.)

  15. Plasma Wall Interaction Phenomena on Tungsten Armour Materials for Fusion Applications

    International Nuclear Information System (INIS)

    Uytdenhouwen, I.; Massaut, V.; Linke, J.; Van Oost, G.

    2008-01-01

    One of the most attractive future complements to present energy sources is nuclear fusion. A large progress was made throughout the last decade from both the physical as the technological area leading to the construction of the ITER machine. One of the key issues that recently received a large interest at international level is focused on the Plasma Wall Interaction (PWI). One of the promising Plasma Facing Materials (PFM) are Tungsten (W) and Tungsten alloys. However, despite the worldwide use and industrial availability of W, the database of physical and mechanical properties is very limited. Especially after fusion relevant neutron irradiation and PWI phenomena, most of the properties are still unknown. The plasma fuel consists out of deuterium (D) and tritium (T). Tritium is radio-active and therefore an issue from the safety point of view. During steady-state plasma operation of future fusion power plants, the PFM need to extract a power density of ∼10-20 MW/m 2 . On top of this heat, transient events will deposit an additional non-negligible amount of energy (Disruptions, Vertical Displacement Events, Edge Localized Modes) during short durations. These severe heat loads cause cracking and even melting of the surface resulting in a reduced lifetime and the creation of dust. A contribution to the understanding of cracking phenomena under the severe thermal loads is described as well as the properties degradation under neutron irradiation. Several W grades were irradiated in the BR2 reactor (SCK.CEN) and the thermal loads were simulated with the electron-beam facility JUDITH (FZJ). Since knowledge should be gained about the Tritium retention in the PFM for safety and licensing reasons, a unique test facility at SCK.CEN is being set-up. The plasmatron VISION-I will simulate steady state plasmas for Tritium retention studies. The formation of surface cracks and dust, the initial porosity, neutron induced traps, re-deposited material - change the Tritium

  16. Plasma Wall Interaction Phenomena on Tungsten Armour Materials for Fusion Applications

    Energy Technology Data Exchange (ETDEWEB)

    Uytdenhouwen, I. [SCK.CEN - The Belgian Nuclear Research Centre, Institute for Nuclear Materials Science, Boeretang 200, 2400 Mol (Belgium); Forschungszentrum Juelich GmbH, EURATOM-association, D-52425 Juelich (Germany); Department of Applied Physics, Ghent University, Rozier 44, 9000 Ghent (Belgium); Massaut, V. [Department of Applied Physics, Ghent University, Rozier 44, 9000 Ghent (Belgium); Linke, J. [Forschungszentrum Juelich GmbH, EURATOM-association, D-52425 Juelich (Germany); Van Oost, G. [Department of Applied Physics, Ghent University, Rozier 44, 9000 Ghent (Belgium)

    2008-07-01

    One of the most attractive future complements to present energy sources is nuclear fusion. A large progress was made throughout the last decade from both the physical as the technological area leading to the construction of the ITER machine. One of the key issues that recently received a large interest at international level is focused on the Plasma Wall Interaction (PWI). One of the promising Plasma Facing Materials (PFM) are Tungsten (W) and Tungsten alloys. However, despite the worldwide use and industrial availability of W, the database of physical and mechanical properties is very limited. Especially after fusion relevant neutron irradiation and PWI phenomena, most of the properties are still unknown. The plasma fuel consists out of deuterium (D) and tritium (T). Tritium is radio-active and therefore an issue from the safety point of view. During steady-state plasma operation of future fusion power plants, the PFM need to extract a power density of {approx}10-20 MW/m{sup 2}. On top of this heat, transient events will deposit an additional non-negligible amount of energy (Disruptions, Vertical Displacement Events, Edge Localized Modes) during short durations. These severe heat loads cause cracking and even melting of the surface resulting in a reduced lifetime and the creation of dust. A contribution to the understanding of cracking phenomena under the severe thermal loads is described as well as the properties degradation under neutron irradiation. Several W grades were irradiated in the BR2 reactor (SCK.CEN) and the thermal loads were simulated with the electron-beam facility JUDITH (FZJ). Since knowledge should be gained about the Tritium retention in the PFM for safety and licensing reasons, a unique test facility at SCK.CEN is being set-up. The plasmatron VISION-I will simulate steady state plasmas for Tritium retention studies. The formation of surface cracks and dust, the initial porosity, neutron induced traps, re-deposited material - change the Tritium

  17. Steady state technologies for tokamak based fusion neutron sources and hybrids

    International Nuclear Information System (INIS)

    Azizov, E.A.; Kuteev, B.V.

    2015-01-01

    Full text of publication follows. The development of demonstration fusion neutron sources for fusion nuclear science activity and hybrid applications has reached the stage of conceptual design on the basis of tokamak device in Russia. The conceptual design of FNS-ST has been completed in details (plasma current 1.5 MA, magnetic field 1.5 T, major radius 0.5 m, aspect ratio 1.67 and auxiliary heating power up to 15 MW) [1, 2]. A comparison of physical plasma parameters and economics for FNS-ST and a conventional tokamak FNS-CT (plasma current 1.5 MA, magnetic field 6.7 T, major radius 2.25 m, aspect ratio 3 and auxiliary heating power up to 30 MW) has been fulfilled [3]. This study suggested the feasibility to reach 1-20 MW of fusion power using these magnetic configuration options. Nevertheless, the efficiency of neutron production Q remains comparable for both due to the beam fusion input. The total ST-economics for the full project including operation and utilization costs is by a factor of 2 better than of CT. Zero [4] and one-dimensional [5] models have been developed and used in this system analysis. The characteristics of plasma confinement, stability and current drive in operation have been confirmed by numerous benchmarking simulations of modern experiments. Scenarios allowing us to reach and maintain steady state operation have been considered and optimized. The results of these studies will be presented. Prospective technical solutions for SSO-technology systems have been evaluated, and the choice of enabling technologies and materials of the basic FNS options has been made. A conceptual design of a thin-wall water cooled vacuum chamber for heat loadings up to 1.5 MW/m 2 has been fulfilled. The chamber consists of 2 mm Be tiles, pre-shaped CuCrZr 1 mm shell and 1 mm of stainless steel shell as a structural material. A concept of double-null divertor for FNS-ST has been offered that is capable to withstand heat fluxes up to 6 MW/m 2 . Lithium dust

  18. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (P fus ∼ 10-100 MW, β N ∼ 2-3, Q p ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  19. Comparison and analysis of 1D/2D/3D neutronics modeling for a fusion reactor

    International Nuclear Information System (INIS)

    Li, J.; Zeng, Q.; Chen, M.; Jiang, J.; Wu, Y.

    2007-01-01

    During the course of analyzing the characteristics for fusion reactors, the refined calculations are needed to confirm that the nuclear design requirements are met. Since the long computational time is consumed, the refined three-dimensional (3D) representation has been used primarily for establishing the baseline reference values, analyzing problems which cannot be reduced by symmetry considerations to lower dimensions, or where a high level of accuracy is desired locally. The two-dimensional (2D) or one-dimensional (1D) description leads itself readily to resolve many problems, such as the studies for the material fraction optimization, or for the blanket size optimization. The purpose of this paper is to find out the differences among different geometric descriptions, which can guide the way to approximate and simplify the computational model. The fusion power reactor named FDS-II was designed as an advanced fusion power reactor to demonstrate and validate the commercialization of fusion power by Institute of Plasma Physics, Chinese Academy of Science. In this contribution, the dual-cooled lithium lead (DLL) blanket of FDS-II was used as a reference for neutronics comparisons and analyses. The geometric descriptions include 1D concentric sphere model, 1D, 2D and 3D cylinder models. The home-developed multi-functional neutronics analysis code system VisualBUS, the Monte Carlo transport code MCNP and nuclear data library HENDL have been used for these analyses. The neutron wall loading distribution, tritium breeding ratio (TBR) and nuclear heat were calculated to evaluate the nuclear performance. The 3D calculation has been used as a comparison reference because it has the least errors in the treatment of geometry. It is suggested that the value of TBR calculated by the 1D approach should be greater than 1.3 to satisfy the practical need of tritium self-sufficiency. The distribution of nuclear heat based on the 2D and 3D models were similar since they all consider

  20. First wall studies of a laser-fusion hybrid reactor design

    International Nuclear Information System (INIS)

    Hovingh, J.

    1976-09-01

    The design of a first wall for a 20 MW thermonuclear power laser fusion hybrid reactor is presented. The 20 mm thick graphite first wall is located 3.5 m from the DT microexplosion with a thermonuclear yield of 10 MJ. Estimates of the energy deposition, temperature, stresses, and material vaporized from the first wall due to the interaction of the x-rays, charged particle debris, and reflected laser light with the graphite are presented, along with a brief description of the analytical methods used for these estimations. Graphite is a viable first wall material for inertially-confined fusion reactors, with lifetimes of a year possible

  1. DT fusion neutron irradiation of BPNL niobium nickel and 316 stainless steel at 1750C

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation at 175 0 C of 17 niobium wires, one niobium foil, 14 316 stainless steel wires, one 316 stainless steel foil, nine nickel wires, and two nickel foils from BPNL is described. The sample position, beam-on time, neutron dose record, and neutron fluence are given

  2. First-wall design limitations for linear magnetic fusion (LMF) reactors

    International Nuclear Information System (INIS)

    Gryczkowski, G.E.; Krakowski, R.A.; Steinhauer, L.C.; Zumdieck, J.

    1978-01-01

    One approach to the endloss problem in linear magnetic fusion (LMF) uses high magnetic field to reduce the required confinement time. This approach is limited by magnet stresses and bremsstrahlung heating of the first wall; the first-wall thermal-pulsing issue is addressed. Pertinent thermophysical parameters are developed in the context of high-field LMF to identify promising first-wall materials, and thermal fatigue experiments relevant to LMF first walls are reviewed. High-flux first-wall concepts are described which include both solid and evaporating first-wall configurations

  3. Lifetime evaluation for thermal fatigue: application at the first wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Merola, M.; Biggio, M.

    1989-01-01

    Thermal fatigue seems to be the most lifetime limiting phenomenon for the first wall of the next generation Tokamak fusion reactors. This work deals with the problem of the thermal fatigue in relation to the lifetime prediction of the fusion reactor first wall. The aim is to compare different lifetime methodologies among them and with experimental results. To fulfil this purpose, it has been necessary to develop a new numerical methodology, called reduced-3D, especially suitable for thermal fatigue problems

  4. 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.

  5. Innovative three-dimensional neutronics analyses directly coupled with cad models of geometrically complex fusion systems

    International Nuclear Information System (INIS)

    Sawan, M.; Wilson, P.; El-Guebaly, L.; Henderson, D.; Sviatoslavsky, G.; Bohm, T.; Kiedrowski, B.; Ibrahim, A.; Smith, B.; Slaybaugh, R.; Tautges, T.

    2007-01-01

    Fusion systems are, in general, geometrically complex requiring detailed three-dimensional (3-D) nuclear analysis. This analysis is required to address tritium self-sufficiency, nuclear heating, radiation damage, shielding, and radiation streaming issues. To facilitate such calculations, we developed an innovative computational tool that is based on the continuous energy Monte Carlo code MCNP and permits the direct use of CAD-based solid models in the ray-tracing. This allows performing the neutronics calculations in a model that preserves the geometrical details without any simplification, eliminates possible human error in modeling the geometry for MCNP, and allows faster design iterations. In addition to improving the work flow for simulating complex 3- D geometries, it allows a richer representation of the geometry compared to the standard 2nd order polynomial representation. This newly developed tool has been successfully tested for a detailed 40 degree sector benchmark of the International Thermonuclear Experimental Reactor (ITER). The calculations included determining the poloidal variation of the neutron wall loading, flux and nuclear heating in the divertor components, nuclear heating in toroidal field coils, and radiation streaming in the mid-plane port. The tool has been applied to perform 3-D nuclear analysis for several fusion designs including the ARIES Compact Stellarator (ARIES-CS), the High Average Power Laser (HAPL) inertial fusion power plant, and ITER first wall/shield (FWS) modules. The ARIES-CS stellarator has a first wall shape and a plasma profile that varies toroidally within each field period compared to the uniform toroidal shape in tokamaks. Such variation cannot be modeled analytically in the standard MCNP code. The impact of the complex helical geometry and the non-uniform blanket and divertor on the overall tritium breeding ratio and total nuclear heating was determined. In addition, we calculated the neutron wall loading variation in

  6. Neutronics design aspects of reference ARIES-I fusion blanket

    International Nuclear Information System (INIS)

    Cheng, E.T.

    1990-12-01

    A SiC composite blanket concept was recently conceived for a deuterium-tritium burning, 1000 MW(e) tokamak fusion reactor design, ARIES-I. SiC composite structural material was chosen due to its very low activation features. High blanket nuclear performance and thermal efficiency, adequate tritium breeding, and a low level of activation are important design requirements for the ARIES-I reactor. The major approaches, other than using SiC as structural material, in meeting these design requirements, are to employ beryllium, the only low activation neutron multiplying material, and isotopically tailored Li 2 ZrO 3 , a tritium breeding material stable at high temperature, as blanket materials. 5 refs., 4 figs., 2 tabs

  7. A review of the behaviour of graphite under the conditions appropriate for protection of the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Birch, M.; Brocklehurst, J.E.

    1987-12-01

    The material used as a first wall protection in fusion reactor systems will be exposed to 14 MeV neutrons from the fusion reaction and suffer surface bombardment by other energetic particles in the plasma. Graphite is a potential candidate for the first wall material. Calculations are performed of the damaging power of 14 MeV neutrons so that existing graphite irradiation data can be utilised. Such data at high irradiation temperatures are reviewed for a wide range of graphite types, characterised by specific examples, and the application of the data to design calculations is discussed. The erosion/corrosion effect of the plasma at the graphite surface is also considered. Limitations in the state of knowledge are identified, and particular areas of further work are recommended. (author)

  8. Neutron analysis of a hybrid system fusion-fission

    International Nuclear Information System (INIS)

    Dorantes C, J. J.; Francois L, J. L.

    2011-11-01

    The use of energy at world level implies the decrease of natural resources, reduction of fossil fuels, in particular, and a high environmental impact. In view of this problem, an alternative is the energy production for nuclear means, because up to now is one of the less polluting energy; however, the nuclear fuel wastes continue being even a problem without being solved. For the above mentioned this work intends the creation of a device that incorporates the combined technologies of fission and nuclear fusion, called Nuclear Hybrid Reactor Fusion-Fission (HRFF). The HRFF has been designed theoretically with base in experimental fusion reactors in different parts of the world like: United States, Russia, Japan, China and United Kingdom, mainly. The hybrid reactor model here studied corresponds at the Compact Nuclear Facility Source (CNFS). The importance of the CNFS resides in its feasibility, simple design, minor size and low cost; uses deuterium-tritium like main source of neutrons, and as fuel can use the spent fuel of conventional nuclear reactors, such as the current light water reactors. Due to the high costs of experimental research, this work consists on simulating in computer a proposed model of CNFS under normal conditions of operation, to modify the arrangement of the used fuel: MOX and IMF, to analyze the obtained results and to give final conclusions. In conclusion, the HRFF can be a versatile system for the management of spent fuel of light water reactors, so much for the possibility of actinides destruction, like for the breeding of fissile material. (Author)

  9. Neutronics requirements for a 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 Consortium , Boltzmannstraße 2, 85748 Garching (Germany); 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 UT-FUS C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

    2015-10-15

    Highlights: • Discussion and specification of neutronic requirements for a DEMO power plant. • TBR uncertainties are reviewed/discussed and design margins are elaborated. • Limits are given for radiation loads to super-conducting magnets and steel structural components. • Available DEMO results are compared to recommended limits and TBR design target. - Abstract: This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.

  10. Fission fragment simulation of fusion neutron radiation effects on bulk mechanical properties

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; Mitchell, J.B.; Guinan, M.W.; Stuart, R.N.; Borg, R.J.

    1976-01-01

    This research demonstrates the feasibility of using homogeneously-generated fission fragments to simulate high-fluence fusion neutron damage in niobium tensile specimens. This technique makes it possible to measure radiation effects on bulk mechanical properties at high damage states, using conveniently short irradiation times. The primary knock-on spectrum for a fusion reactor is very similar to that produced by fission fragments, and nearly the same ratio of gas atoms to displaced atoms is produced in niobium. The damage from fission fragments is compared to that from fusion neutrons and fission reactor neutrons in terms of experimentally measured yield strength increase, transmission electron microscopy (TEM) observations, and calculated damage energies

  11. Lithium adsorption by the first wall of fusion reactor-tokamak

    International Nuclear Information System (INIS)

    Bakunin, O.G.

    1989-01-01

    Lithium adsorption by the first wall of fusion reactor under stationary conditions and in the absence of chemical reactions is considered. Possibility of achieving 70% coating of the wall with lithium which can lead to sufficient decrease of sputtering is shown. 5 refs.; 5 figs

  12. Lifetime estimates of a fusion reactor first wall by linear damage summation and strain range partitioning methods

    International Nuclear Information System (INIS)

    Liu, K.C.; Grossbeck, M.L.

    1979-01-01

    A generalized model of a first wall made of 20% cold-worked steel was examined for neutron wall loadings ranging from 2 to 5 MW/m 2 . A spectrum of simplified on-off duty cycles was assumed with a 95% burn time. Independent evaluations of cyclic lifetimes were based on two methods: the method of linear damage summation currently being employed for use in ASME high-temperature design Code Case N-47 and that of strain range partitioning being studied for inclusion in the design code. An important point is that the latter method can incorporate a known decrease in ductility for materials subject to irradiation as a parameter, so low-cycle fatigue behavior can be estimated for irradiated material. Lifetimes predicted by the two methods agree reasonably well despite their diversity in concept. Lack of high-cycle fatigue data for the material tested at temperatures within the range of our interest precludes making conclusions on the accuracy of the predicted results, but such data are forthcoming. The analysis includes stress relaxation due to thermal and irradiation-induced creep. Reduced ductility values from irradiations that simulate the environment of the first wall of a fusion reactor were used to estimate the lifetime of the first wall under irradiation. These results indicate that 20% cold-worked type 316 stainless steel could be used as a first-wall material meeting a 8 to 10 MW-year/m 2 lifetime goal for a neutron wall loading of about 2 MW-year/m 2 and a maximum temperature of about 500 0 C

  13. The EUROBALL neutron wall - design and performance tests of neutron detectors

    CERN Document Server

    Skeppstedt, Ö; Lindström, L; Wadsworth, R; Hibbert, I; Kelsall, N; Jenkins, D; Grawe, H; aGórska, M; Moszynski, M; Sujkowski, Z; Wolski, D; Kapusta, M; Hellström, M; Kalogeropoulos, S; Oner, D; Johnson, A; Cederkäll, J; Klamra, W; Nyberg, J; Weiszflog, M; Kay, J; Griffiths, R; Garces-Narro, J; Pearson, C; Eberth, J

    1999-01-01

    The mechanical design of the EUROBALL neutron wall and neutron detectors, and their performance measured with a sup 2 sup 4 sup 6 sup , sup 2 sup 4 sup 8 Cm fission source are described. The array consists of 15 pseudohexaconical detector units subdivided into three, 149 mm high, hermetically separated segments and a smaller central pentagonal unit subdivided into five segments. The detectors are filled with Bicron BC501A liquid scintillator. Each section of the hexaconical detectors is viewed by a 130 mm diameter Philips XP4512PA photomultiplier while the sections of pentagonal detectors are viewed by Philips XP4312B PMTs. The tests of n-gamma discrimination performed by zero-crossing and time-of-flight methods show a full separation of gamma- and neutron events down to 50 keV recoil electron energy. These tests demonstrate the excellent timing properties of the detectors and an average time resolution of 1.56 ns. The factors determining the efficiency of neutron detectors are discussed. The total efficiency...

  14. Proceedings of a specialists' meeting on neutron activation cross sections for fission and fusion energy applications

    International Nuclear Information System (INIS)

    Wagner, M.; Vonach, H.

    1990-01-01

    These proceedings of a specialists' meeting on neutron activation cross sections for fission and fusion energy applications are divided into 4 sessions bearing on: - data needs: 4 conferences - experimental work: 11 conferences - theoretical work: 4 conferences - evaluation work: 5 conferences

  15. Surface erosion of fusion reactor components due to radiation blistering and neutron sputtering

    International Nuclear Information System (INIS)

    Das, S.K.; Kaminsky, M.

    1975-01-01

    Radiation blistering and neutron sputtering can lead to the surface erosion of fusion reactor components exposed to plasma radiations. Recent studies of methods to reduce the surface erosion caused by these processes are discussed

  16. First wall thermal hydraulic models for fusion blankets

    International Nuclear Information System (INIS)

    Fillo, J.A.

    1980-01-01

    Subject to normal and off-normal reactor conditions, thermal hydraulic models of first walls, e.g., a thermal mass barrier, a tubular shield, and a radiating liner are reviewed. Under normal operation the plasma behaves as expected in a predicted way for transient and steady-state conditions. The most severe thermal loading on the first wall occurs when the plasma becomes unstable and dumps its energy on the wall in a very short period of time (milliseconds). Depending on the plasma dump time and area over which the energy is deposited may result in melting of the first wall surface, and if the temperature is high enough, vaporization

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

    International Nuclear Information System (INIS)

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

    1981-12-01

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

  18. Design of intense neutron source for fusion material study and the role of universities

    International Nuclear Information System (INIS)

    Ishino, Shiori

    1993-01-01

    Need and requirement for the intense neutron source for fusion materials study have been discussed for many years. Recently, international climate has been becoming gradually maturing to consider this problem more seriously because of the recognition of crucial importance of solving materials problems for fusion energy development. The present symposium was designed to discuss the problems associated with the intense neutron source for material irradiation studies which will have a potential for the National Institute for Fusion Science to become one of the important future research areas. The symposium comprises five sessions; first, the role of materials research in fusion development strategies was discussed followed by a brief summary of current IFMIF (International Fusion Materials Irradiation Facility) activity. Despite the pressing need for intense fusion neutron source, currently available neutron sources are reactor or accelerator based sources of which FFTF and LASREF were discussed. Then, various concepts of intense neutron source candidates were presented including ESNIT, which are currently under design by JAERI. In the fourth session, discussions were made on the study of materials with the intense neutron source from the viewpoint of materials scientists and engineers as the user of the facility. This is followed by discussions on the role of universities from the two stand points, namely, fusion irradiation studies and fusion materials development. Finally summary discussions were made by the participants, indicating important role fundamental studies in universities for the full utilization of irradiation data and the need of pure 14 MeV neutron source for fundamental studies together with the intense surrogate neutron sources. (author)

  19. Correlation of DT and DD fusion neutron damage in silicon surface barrier detector

    International Nuclear Information System (INIS)

    Iida, Toshiyuki; Sueyoshi, Yasuhiro; Sunarno; Takahashi, Akito

    1994-01-01

    In order to examine the correlation of DT and DD fusion neutron damage in Si, a silicon surface barrier detector (Si-SBD) was irradiation with neutrons from a deuteron accelerator. The leakage current increased proportionally with neutron fluence, which determined the neutron damage constant for the Si-SBD. The correlation factor of the DT and DD neutron damage in the Si-SBD was determined from the ratio of the DT and DD neutron damage constants and was found to be 2.3. We also calculated the rate of DT and DD neutron displacement damage for Si by using the TRIM-90 computer program and actual data on neutron reactions in the Si-SBD. The correlation factor of DT and DD neutron damage from the calculation agreed with that from the Si-SBD irradiation experiment. (author)

  20. Fusion Neutronic Source deuterium endash tritium neutron spectrum measurements using natural diamond detectors

    International Nuclear Information System (INIS)

    Krasilnikov, A.V.; Kaneko, J.; Isobe, M.; Maekawa, F.; Nishitani, T.

    1997-01-01

    Two natural diamond detectors (NDDs) operating at room temperature were used for Fusion Neutronics Source (FNS) deuterium endash tritium (DT) neutron spectra measurements at different points around the tritium target and for different deuteron beam energies. Energy resolution of both NDDs were measured, with values 1.95% and 2.8%. Due to the higher energy resolution of one of the two NDDs studied it was possible to measure the shape of the DT neutron energy distribution and its broadening due to deuteron scattering inside the target. The influence of pulse pileup on the energy resolution of the combined system (NDD+electronics) at count rates up to 3.8x10 5 counts/s was investigated. A 3.58% energy resolution for the spectrometric system based on NDD and a 0.25 μs shaping time amplifier has been measured at a count rate of 5.7x10 5 counts/s. It is shown that special development of a fast pulse signal processor is necessary for NDD based spectrometry at count rates of approximately 10 6 counts/s. copyright 1997 American Institute of Physics

  1. Direct integration multiple collision integral transport analysis method for high energy fusion neutronics

    International Nuclear Information System (INIS)

    Koch, K.R.

    1985-01-01

    A new analysis method specially suited for the inherent difficulties of fusion neutronics was developed to provide detailed studies of the fusion neutron transport physics. These studies should provide a better understanding of the limitations and accuracies of typical fusion neutronics calculations. The new analysis method is based on the direct integration of the integral form of the neutron transport equation and employs a continuous energy formulation with the exact treatment of the energy angle kinematics of the scattering process. In addition, the overall solution is analyzed in terms of uncollided, once-collided, and multi-collided solution components based on a multiple collision treatment. Furthermore, the numerical evaluations of integrals use quadrature schemes that are based on the actual dependencies exhibited in the integrands. The new DITRAN computer code was developed on the Cyber 205 vector supercomputer to implement this direct integration multiple-collision fusion neutronics analysis. Three representative fusion reactor models were devised and the solutions to these problems were studied to provide suitable choices for the numerical quadrature orders as well as the discretized solution grid and to understand the limitations of the new analysis method. As further verification and as a first step in assessing the accuracy of existing fusion-neutronics calculations, solutions obtained using the new analysis method were compared to typical multigroup discrete ordinates calculations

  2. Laser fusion reactor design in a fast ignition with a dry wall chamber

    International Nuclear Information System (INIS)

    Ogawa, Yichi; Goto, Takuya; Ninomiya, Daisuke; Hiwatari, Ryoji; Asaoka, Yoshiyuki; Okano, Kunihiko

    2007-01-01

    One of the critical issues in laser fusion reactor design is high pulse heat load on the first wall by the X-rays and the fast/debris ions from fusion burn. There are mainly two concepts for the first wall of laser fusion reactor, a dry wall and a liquid metal wall. We should notice that the fast ignition method can achieve sufficiently high pellet gain with smaller (about 1/10 of the conventional central ignition method) input energy. To take advantage of this property, the design of a laser fusion reactor with a small size dry wall chamber may become possible. Since a small fusion pulse leads to a small electric power, high repetition of laser irradiation is required to keep sufficient electric power. Then we tried to design a laser fusion reactor with a dry wall chamber and a high repetition laser. This is a new challenging path to realize a laser fusion plant. Based on the point model of the core plasma, we have estimated that fusion energy in one pulse can be reduced to be 40 MJ with a pellet gain around G>100. To evaluate the validity of this simple estimation and to optimize the pellet design and the pulse shaping for the fast ignition scenario, we have introduced 1-D hydrodynamic simulation code ILESTA-1D and carried out implosion simulations. Since the code is one-dimensional, the detailed physics process of fast heating cannot be reproduced. Thus the fast heating is reflected in the code as the additional artificial heating source in the energy equation. It is modeled as a homogeneous heating of electrons in core region at the time just before when the maximum compression is achieved. At present we obtained the pellet gain G∝100 with the same input energy as the above estimation by a simple point model (350kJ for implosion, 50kJ for heating and assuming 20% coupling of heating laser). A dry wall is exposed to several threats due to the cyclic load by the high energy X-ray and charged particles: surface melting, physical and chemical sputtering

  3. Neutron imaging for inertial confinement fusion and molecular optic imaging

    International Nuclear Information System (INIS)

    Delage, O.

    2010-01-01

    Scientific domains that require imaging of micrometric/nano-metric objects are dramatically increasing (Plasma Physics, Astrophysics, Biotechnology, Earth Sciences...). Difficulties encountered in imaging smaller and smaller objects make this research area more and more challenging and in constant evolution. The two scientific domains, through which this study has been led, are the neutron imaging in the context of the inertial confinement fusion and the fluorescence molecular imaging. Work presented in this thesis has two main objectives. The first one is to describe the instrumentation characteristics that require such imagery and, relatively to the scientific domains considered, identify parameters likely to optimize the imaging system accuracy. The second one is to present the developed data analysis and reconstruction methods able to provide spatial resolution adapted to the size of the observed object. Similarities of numerical algorithms used in these two scientific domains, which goals are quiet different, show how micrometric/nano-metric object imaging is a research area at the border of a large number of scientific disciplines. (author)

  4. Fusion neutron effects on magnetoresistivity of copper stabilizer materials

    International Nuclear Information System (INIS)

    Guinan, M.W.; Van Konynenburg, R.A.

    1983-01-01

    Eight copper wires were repeatedly irradiated at 4.2 to 4.4 K with 14.8 MV neutrons and isochronally annealed at temperatures up to 34 0 C for a total of five cycles. Their electrical resistances were monitored during irradiation under zero applied magnetic field. After each irradiation the magnetoresistances were measured in applied transverse magnetic fields of up to 12 T. Then the samples were isochronally annealed to observe the recovery of the resistivity and magnetoresistivity. After each anneal at the highest temperature (34 0 C), some of the damage remained and contributed to the damage state observed following the subsequent irradiation. In this way, we were able to observe how the changes in magnetoresistance would accumulate during the repeated irradiations and anneals expected to be characteristic of fusion reactor magnets. For each succeeding irradiation the fluence was chosen to produce approximately the same final magnetoresistance at 12 T, taking account of the accumulating residual radiation damage. The increment of magnetoresistivity added by the irradiation varied from 35 to 65% at 12 T and from 50 to 90% at 8 T for the various samples

  5. Neutron spectrometer for DD/DT burning ratio measurement in fusion experimental reactor

    International Nuclear Information System (INIS)

    Asai, Keisuke; Naoi, Norihiro; Iguchi, Tetsuo; Watanabe, Kenichi; Kawarabayashi, Jun; Nishitani, Takeo

    2006-01-01

    The most feasible fuels for a fusion reactor are D (Deuterium) and T (Tritium). DD and/or DT fusion reaction or nuclear burning reaction provides two kinds of neutrons, DD neutron and DT neutron, respectively. DD/DT burning ratio, which can be estimated by DD/DT neutron ratio in the burning plasma, is essential for burn control, alpha particle emission rate monitoring and tritium fuel cycle estimation. Here we propose a new neutron spectrometer for the absolute DD/DT burning ratio measurement. The system consists of a Proton Recoil Telescope (PRT) and a Time-of-Flight (TOF) technique. We have conducted preliminary experiments with a prototype detector and a DT neutron beam (φ20 mm) at the Fusion Neutronics Source, Japan Atomic Energy Agency (JAEA), to assess its basic performance. The detection efficiency obtained by the experiment is consistent with the calculation results in PRT, and sufficient energy resolution for the DD/DT neutron discrimination has been achieved in PRT and TOF. The validity of the Monte Carlo calculation has also been confirmed by comparing the experimental results with the calculation results. The design consideration of this system for use in ITER (International Thermonuclear Experimental Reactor) has shown that this system is capable of monitoring the line-integrated DD/DT burning ratio for the plasma core line of sight with the required measurement accuracy of 20% in the upper 4 decades of the ITER operation (fusion power: 100 kW-700 MW). (author)

  6. Status of neutron dosimetry and damage analysis for the fusion materials program

    International Nuclear Information System (INIS)

    Greenwood, L.R.

    1979-01-01

    The status of neutron flux and spectral measurements is described for fusion material irradiations at reactor, T(d,n), Be(d,n), and spallation neutron sources. Such measurements are required for the characterization of an irradiation in terms of displacement damage, gas and transmutant production. Emphasis is placed on nuclear data deficiencies with specific recommendations for cross section measurements and calculations

  7. Chamber and Wall Response to Target Implosion in Inertial and Z-Pinch Fusion and Lithography Devices

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.; Morozov, V.; Sizyuk, V.

    2006-01-01

    The chamber walls, both solid and liquid, in inertial fusion energy (IFE) and Z-pinch reactors and Lithography devices are exposed to harsh conditions following each target implosion or pinching of plasma. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to target implosion. Detailed models have been developed for reflected laser light, emitted photons, neutrons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The hydrodynamic response of chamber walls in bare or in gas-filled cavities and the photon transport of the deposited energy has been calculated by means of new and advanced numerical techniques for accurate shock treatment and propagation. These models include detail media hydrodynamics, non-LTE multi-group for both continuum and line radiation transport, and dynamics of eroded debris resulting from the intense energy deposition. The focus of this study is to critically assess the reliability and the dynamic response of chamber walls in various proposed protection methods for IFE systems. Key requirements are that: (i) the chamber wall accommodates the cyclic energy deposition while providing the required lifetime due to various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls, and (ii) after each shot the chamber is cleared and returned to a quiescent state in preparation for the target injection and the firing of the driver for the subsequent shot. This paper investigates in details these two important issues and found that the required operating frequency of the IFE reactors for power production may be severely limited due to these two requirements. (author)

  8. Recent developments in fusion first wall, blanket, and shield technology

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1983-01-01

    This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

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

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  11. Can wall and limiter erosion be eliminated in fusion reactors

    International Nuclear Information System (INIS)

    Norem, J.H.

    1981-10-01

    A pump limiter system is described which is compatible with in-situ recoating of the limiter surface. The recoating could be done during normal tokamak operation. We have shown how this system is compatible with most of the constraints of fusion reactor operation and might provide a significant advantage over magnetic diverter and some other pump limiter geometries

  12. Tensile property changes of metals and irradiated to low doses with fission, fusion and spallation neutrons

    International Nuclear Information System (INIS)

    Heinisch, H.L.; Hamilton, M.L.; Sommer, W.F.; Ferguson, P.D.

    1992-01-01

    The objective of this work is to investigate the effects of the neutron energy spectrum in low dose irradiations on the microstructures and mechanical properties of metals. Radiation effects due to low doses of spallation neutrons are compared directly to those produced by fission and fusion neutrons. Yield stress changes of pure Cu, alumina-dispersion-strengthened Cu and AISI 316 stainless steel irradiated at 36-55 C in the Los Alamos Spallation Radiation Effects Facility (LASREF) are compared with earlier results of irradiations at 90 C using 14 MeV D-T fusion neutrons at the Rotating Target Neutron Source and fission reactor neutrons in the Omega West Reactor. At doses up to 0.04 displacements per atom (dpa), the yield stress changes due to the three quite different neutron spectra correlate well on the basis of dpa in the stainless steel and the Cu alloy. However, in pure Cu, the measured yield stress changes due to spallation neutrons were anomalously small and should be verified by additional irradiations. With the exception of pure Cu, the low dose, low temperature experiments reveal no fundamental differences in radiation hardening by fission, fusion or spallation neutrons when compared on the basis of dpa

  13. Simulation of fusion first-wall environment in a fission reactor

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Kulcinski, G.L.; Longhurst, G.R.

    1982-01-01

    A novel concept to produce a realistic simulation of a fusion first-wall test environment has been proposed recently. This concept takes advantage of the (/eta/, α) reaction in 59 Ni to produce a high internal helium content in the metal while using the 3 He (/eta/, /rho/)T reaction in the gas surrounding the specimen to produce an external heat and particle flux. Models to calculate heat flux, erosion rate, implantation, and damage rate to the walls of the test module are presented. Preliminary results show that a number of important fusion technology issues could be tested experimentally in a fission reactor such as the Engineering Test Reactor

  14. Neutronics analysis for aqueous self-cooled fusion reactor blankets

    International Nuclear Information System (INIS)

    Varsamis, G.; Embrechts, M.J.; Jaffa, R.; Steiner, D.; Deutsch, L.; Gierszewski, P.

    1986-06-01

    The tritium breeding performance of several Aqueous Self-Cooled Blanket (ASCB) configurations for fusion reactors has been evaluated. The ASCB concept employs small amounts of lithium compound dissolved in light or heavy water to serve as both coolant and breeding medium. The inherent simplicity of this concept allows the development of blankets with minimal technological risk. The tritium breeding performance of the ASCB concept is a critical issue for this family of blankets. Contrary to conventional blanket designs there will be a significant contribution to the tritium breeding ratio (TBR) in the water coolant/breeder of duct shields, and the 3-D TBR will therefore be similar to the 1-D TBR. The tritium breeding performance of an ASCB for a MARS-like-tandem reactor and an ASCB based breeding-shield for the Next European Torus (NET) are assessed. Two design options for the MARS-like blanket are discussed. One design employs a vanadium first wall, and zircaloy for the structural material. The trade-offs between light water and heavy water cooling options for this zircaloy blanket are discussed. The second design option for MARS relies on the use of a vanadium alloy as the stuctural material, and heavy water as the coolant. It is demonstrated that both design options lead to low-activation blankets that allow class C burial. The breeder-shield for NET consists of a water-cooled stainless steel shield

  15. A D-D/D-T fusion reaction based neutron generator system for liver tumor BNCT

    International Nuclear Information System (INIS)

    Koivunoro, H.; Lou, T.P.; Leung, K. N.; Reijonen, J.

    2003-01-01

    Boron-neutron capture therapy (BNCT) is an experimental radiation treatment modality used for highly malignant tumor treatments. Prior to irradiation with low energetic neutrons, a 10B compound is located selectively in the tumor cells. The effect of the treatment is based on the high LET radiation released in the 10 B(n,α) 7 Li reaction with thermal neutrons. BNCT has been used experimentally for brain tumor and melanoma treatments. Lately applications of other severe tumor type treatments have been introduced. Results have shown that liver tumors can also be treated by BNCT. At Lawrence Berkeley National Laboratory, various compact neutron generators based on D-D or D-T fusion reactions are being developed. The earlier theoretical studies of the D-D or D-T fusion reaction based neutron generators have shown that the optimal moderator and reflector configuration for brain tumor BNCT can be created. In this work, the applicability of 2.5 MeV neutrons for liver tumor BNCT application was studied. The optimal neutron energy for external liver treatments is not known. Neutron beams of different energies (1eV < E < 100 keV) were simulated and the dose distribution in the liver was calculated with the MCNP simulation code. In order to obtain the optimal neutron energy spectrum with the D-D neutrons, various moderator designs were performed using MCNP simulations. In this article the neutron spectrum and the optimized beam shaping assembly for liver tumor treatments is presented

  16. An accelerated beam-plasma neutron/proton source and early application of a fusion plasma

    International Nuclear Information System (INIS)

    Ohnishi, M.; Yoshikawa, K.; Yamamoto, Y.; Hoshino, C.; Masuda, K.; Miley, G.; Jurczyk, B.; Stubbers, R.; Gu, Y.

    1999-01-01

    We measured the number of the neutrons and protons produced by D-D reactions in an accelerated beam-plasma fusion and curried out the numerical simulations. The linear dependence of the neutron yield on a discharge current indicates that the fusion reactions occur between the background gas and the fast particles. i.e. charge exchanged neutrals and accelerated ions. The neutron yield divided by (fusion cross section x ion current x neutral gas pressure) still possesses the dependence of the 1.2 power of discharge voltage. which shows the fusion reactions are affected by the electrostatic potential built-up in the center. The measured proton birth profiles suggest the existence of a double potential well, which is supported by the numerical simulations. (author)

  17. Method and apparatus to produce and maintain a thick, flowing, liquid lithium first wall for toroidal magnetic confinement DT fusion reactors

    Science.gov (United States)

    Woolley, Robert D.

    2002-01-01

    A system for forming a thick flowing liquid metal, in this case lithium, layer on the inside wall of a toroid containing the plasma of a deuterium-tritium fusion reactor. The presence of the liquid metal layer or first wall serves to prevent neutron damage to the walls of the toroid. A poloidal current in the liquid metal layer is oriented so that it flows in the same direction as the current in a series of external magnets used to confine the plasma. This current alignment results in the liquid metal being forced against the wall of the toroid. After the liquid metal exits the toroid it is pumped to a heat extraction and power conversion device prior to being reentering the toroid.

  18. First wall and blanket stresses induced by cyclic fusion core operations

    International Nuclear Information System (INIS)

    Bohachevsky, I.O.; Kostoff, R.N.

    1981-01-01

    An analysis is made of cyclic thermal loads and stresses for the complete range of operating conditions. Two critical components were examined; the solid wall adjacent to the fusion plasma (first wall) and the fuel elements in the high power density region of the blanket. Simple closed form expressions were derived for temperature increases and thermal stresses that may be evaluated conveniently and rapidly and the values compared for different systems

  19. A three-bar model for ratcheting of fusion reactor first wall

    International Nuclear Information System (INIS)

    Wolters, J.; Majumdar, S.

    1994-12-01

    First wall structures of fusion reactors are subjected to cyclic bending stresses caused by inhomogeneous temperature distribution during plasma burn cycles and by electromagnetically induced impact loads during plasma disruptions. Such a combination of loading can potentially lead to ratcheting or incremental accumulation of plastic strain with cycles. An elastic-plastic three-bar model is developed to investigate the ratcheting behavior of the first wall

  20. Data fusion of ultrasound and GPR signals for analysis of historic walls

    International Nuclear Information System (INIS)

    Salazar, A; Gosalbez, J; Safont, G; Vergara, L

    2012-01-01

    This paper presents an application of ultrasounds and ground-penetrating radar (GPR) for analysis of historic walls. The objectives are to characterize the deformation of a historic wall under different levels of load weights and to obtain an enhanced image of the wall. A new method that fuses data from ultrasound and GPR traces is proposed which is based on order statistics digital filters. Application results are presented for non destructive testing (NDT) of two replicates of historic ashlars' masonry walls: the first one homogeneous and the second one containing controlled defects such as cracks and nooks. The walls are measured separately using ultrasounds and GPR at different load steps. Time and frequency parameters extracted from the signals and different B-Scans for each of the NDT techniques are obtained. After this, a new fused representation is obtained, which results demonstrate the improvement of characterization and defect detection in historic walls using data fusion.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  2. Prospects for fusion applications of reversed-field pinches

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. Transmutation and activation of fusion reactor wall and structural materials

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1979-01-01

    This report details the extent of the nuclear data needed for inclusion in a data library to be used for general assessments of fusion reactor structure activation and transmutation, describes the sources of data available, reviews the literature and explores the reliability of current calculations by providing an independent assessment of the activity inventory to be expected from five structural materials in a simple blanket design for comparison with the results of other workers. An indication of the nuclear reactions which make important contributions to the activity, transmutation and gas production rates for these structural materials is also presented. (author)

  4. SIMS investigations of wall coatings for application in nuclear fusion reactors

    International Nuclear Information System (INIS)

    Friedbacher, G.; Virag, A.; Grasserbauer, M.; Esser, H.G.; Wienhold, P.

    1989-01-01

    Carbon coated metals have proven to be useful materials for limiters and the first wall in fusion reactors. In this paper SIMS investigations of a-C:B single coated and a-C:D/a-C:B double coated stainless steel samples, which have been exposed to TOKAMAK discharges in deuterium and helium, are described. (orig.)

  5. First-wall and blanket engineering development for magnetic-fusion reactors

    International Nuclear Information System (INIS)

    Baker, C.; Herman, H.; Maroni, V.; Turner, L.; Clemmer, R.; Finn, P.; Johnson, C.; Abdou, M.

    1981-01-01

    A number of programs in the USA concerned with materials and engineering development of the first wall and breeder blanket systems for magnetic-fusion power reactors are described. Argonne National Laboratory has the lead or coordinating role, with many major elements of the research and engineering tests carried out by a number of organizations including industry and other national laboratories

  6. NATO Advanced Study Institute entitled Physics of Plasma-Wall Interactions in Controlled Fusion

    CERN Document Server

    Behrisch, R; Physics of plasma-wall interactions in controlled fusion

    1986-01-01

    Controlled thermonuclear fusion is one of the possible candidates for long term energy sources which will be indispensable for our highly technological society. However, the physics and technology of controlled fusion are extremely complex and still require a great deal of research and development before fusion can be a practical energy source. For producing energy via controlled fusion a deuterium-tritium gas has to be heated to temperatures of a few 100 Million °c corres­ ponding to about 10 keV. For net energy gain, this hot plasma has to be confined at a certain density for a certain time One pro­ mising scheme to confine such a plasma is the use of i~tense mag­ netic fields. However, the plasma diffuses out of the confining magnetic surfaces and impinges on the surrounding vessel walls which isolate the plasma from the surrounding air. Because of this plasma wall interaction, particles from the plasma are lost to the walls by implantation and are partially reemitted into the plasma. In addition, wall...

  7. Spectrum shaping assessment of accelerator-based fusion neutron sources to be used in BNCT treatment

    Science.gov (United States)

    Cerullo, N.; Esposito, J.; Daquino, G. G.

    2004-01-01

    Monte Carlo modelling of an irradiation facility, for boron neutron capture therapy (BNCT) application, using a set of advanced type, accelerator based, 3H(d,n) 4He (D-T) fusion neutron source device is presented. Some general issues concerning the design of a proper irradiation beam shaping assembly, based on very hard energy neutron source spectrum, are reviewed. The facility here proposed, which represents an interesting solution compared to the much more investigated Li or Be based accelerator driven neutron source could fulfil all the medical and safety requirements to be used by an hospital environment.

  8. Review of melting and evaporation of fusion-reactor first walls

    International Nuclear Information System (INIS)

    Fillo, J.A.; Makowitz, H.

    1981-01-01

    The most severe thermal loading on the first wall will occur when the plasma becomes unstable resulting in a hard plasma disruption or at the end of a discharge when the plasma is dumped on the wall in a very short period of time. Hard plasma disruptions are of particular concern in future fusion reactors where the thermal energy of the plasma may reach values on the order of 300 MJ. Sufficiently high heating rates can occur to melt the first wall surface, and the temperature can increase resulting in vaporization. Thermal models are reviewed which treat these problems

  9. A Charged Particle Veto Wall for the Large Area Neutron Array (LANA)

    Science.gov (United States)

    Zhu, K.; Chajecki, Z.; Anderson, C.; Bromell, J.; Brown, K.; Crosby, J.; Kodali, S.; Lynch, W. G.; Morfouace, P.; Sweany, S.; Tsang, M. B.; Tsang, C.; Brett, J. J.; Swaim, J. L.

    2017-09-01

    Comparison of neutrons and protons emitted in heavy ion collisions is one of the observables to probe symmetry energy, which is related to the properties of neutron star. In general, neutrons are difficult to measure and neutron detectors are not as easy to use or as widely available as charged particle detectors. Two neutron walls (NW) called LANA exist at the National Superconducting Cyclotron Laboratory. Although the NSCL NW attains excellent discrimination of γ rays and neutron, it fails to discriminate charged particles from neutrons. To ensure near 100% rejection of charged particles, a Charged Particle Veto Wall (VW) is being jointly built by Michigan State University and Western Michigan University. It will be placed in front of one NW. To increase efficiency in detecting neutrons, the second neutron wall is stacked behind it. In this presentation, I will discuss the design, construction and testing of the VW together with the LANA in preparation of two approved NSCL experiments to probe the density and momentum dependence of the symmetry energy potentials in the equation state of the asymmetric nuclear matter. This material is based upon work supported by the National Science Foundation under Grant No. PHY 1565546.

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

    International Nuclear Information System (INIS)

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

    1997-10-01

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

  11. The high-density Z-pinch as a pulsed fusion neutron source for fusion nuclear technology and materials testing

    International Nuclear Information System (INIS)

    Krakowski, R.A.; Sethian, J.D.; Hagenson, R.L.

    1989-01-01

    The dense Z-pinch (DZP) is one of the earliest and simplest plasma heating and confinement schemes. Recent experimental advances based on plasma initiation from hair-like (10s μm in radius) solid hydrogen filaments have so far not encountered the usually devastating MHD instabilities that plagued early DZP experiments. These encouraging results along with debt of a number of proof-of principle, high-current (1--2 MA in 10--100 ns) experiments have prompted consideration of the DZP as a pulsed source of DT fusion neutrons of sufficient strength (/dot S//sub N/ ≥ 10 19 n/s) to provide uncollided neutron fluxes in excess of I/sub ω/ = 5--10 MW/m 2 over test volumes of 10--30 litre or greater. While this neutron source would be pulsed (100s ns pulse widths, 10--100 Hz pulse rate), giving flux time compressions in the range 10 5 --10 6 , its simplicity, near-time feasibility, low cost, high-Q operation, and relevance to fusion systems that may provide a pulsed commercial end-product (e.g., inertial confinement or the DZP itself) together create the impetus for preliminary considerations as a neutron source for fusion nuclear technology and materials testings. The results of a preliminary parametric systems study (focusing primarily on physics issues), conceptual design, and cost versus performance analyses are presented. The DZP promises an expensive and efficient means to provide pulsed DT neutrons at an average rate in excess of 10 19 n/s, with neutron currents I/sub ω/ /approx lt/ 10 MW/m 2 over volumes V/sub exp/ ≥ 30 litre using single-pulse technologies that differ little from those being used in present-day experiments. 34 refs., 17 figs., 6 tabs

  12. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    Science.gov (United States)

    Velarde, G.; Perlado, J. M.; Alonso, E.; Alonso, M.; Domínguez, E.; Rubiano, J. G.; Gil, J. M.; Gómez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martínez-Val, J. M.; Mínguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.

    2001-05-01

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT x fuel with a small tritium initial content ( x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures (≫100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower Te and to enhance radiation losses, reducing the plasma temperature, Ti. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination barriers in Si

  13. Advances in implosion physics, alternative targets design, and neutron effects on heavy ion fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velarde, G.; Perlado, J.M. E-mail: mperlado@denim.upm.es; Alonso, E.; Alonso, M.; Dominguez, E.; Rubiano, J.G.; Gil, J.M.; Gomez del Rio, J.; Lodi, D.; Malerba, L.; Marian, J.; Martel, P.; Martinez-Val, J.M.; Minguez, E.; Piera, M.; Ogando, F.; Reyes, S.; Salvador, M.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P

    2001-05-21

    The coupling of a new radiation transport (RT) solver with an existing multimaterial fluid dynamics code (ARWEN) using Adaptive Mesh Refinement named DAFNE, has been completed. In addition, improvements were made to ARWEN in order to work properly with the RT code, and to make it user-friendlier, including new treatment of Equations of State, and graphical tools for visualization. The evaluation of the code has been performed, comparing it with other existing RT codes (including the one used in DAFNE, but in the single-grid version). These comparisons consist in problems with real input parameters (mainly opacities and geometry parameters). Important advances in Atomic Physics, Opacity calculations and NLTE atomic physics calculations, with participation in significant experiments in this area, have been obtained. Early published calculations showed that a DT{sub x} fuel with a small tritium initial content (x<3%) could work in a catalytic regime in Inertial Fusion Targets, at very high burning temperatures ({>=}100 keV). Otherwise, the cross-section of DT remains much higher than that of DD and no internal breeding of tritium can take place. Improvements in the calculation model allow to properly simulate the effect of inverse Compton scattering which tends to lower T{sub e} and to enhance radiation losses, reducing the plasma temperature, T{sub i}. The neutron activation of all natural elements in First Structural Wall (FSW) component of an Inertial Fusion Energy (IFE) reactor for waste management, and the analysis of activation of target debris in NIF-type facilities has been completed. Using an original efficient modeling for pulse activation, the FSW behavior in inertial fusion has been studied. A radiological dose library coupled to the ACAB code is being generated for assessing impact of environmental releases, and atmospheric dispersion analysis from HIF reactors indicate the uncertainty in tritium release parameters. The first recognition of recombination

  14. Investigation of neutron emission in a cold fusion experiment in palladium

    International Nuclear Information System (INIS)

    Szustakowski, M.; Farny, J.; Muniak, M.; Nowak, A.; Parys, P.; Skrzeczanowski, W.; Socha, R.; Teter, J.; Wolski, J.; Wolowski, J.; Woryna, E.

    1989-01-01

    This paper reports on the experiments dealing with performance of nuclear fusion at room temperature actually which create a great sensation and are carried out in various laboratories. This interest arises from the results achieved by Fleischmann and Pons, and it results from their paper that there exists a possibility of obtaining an ignition owing to nuclear fusion reactions during usual electrochemical process--namely the electrolysis of D O with use of the system of Pd-Pt electrodes. From this reason the measurements of the yield and behavior of neutron emission give the information about processes of interest. At the IPPLM the cold fusion experiments have been conducted from the beginning of April 1989. In the first experiment the reliable evidence of neutron emission was obtained. A number of irregularly repeated neutron pulses of the level of 10 5 per pulse was recorded. The measurements of the neutron emission, in this experiment, were performed with the use of three independent methods employing the 2.5 MeV neutron spectrometer, the scintillation neutron detector as well as the nuclear track detector. neutron emission had been first recorded after 106 hours of the electrolysis process of D 2 O

  15. Effect of fusion burn cycle on first wall swelling

    International Nuclear Information System (INIS)

    Choi, Y.H.; Bement, A.L.; Russell, K.C.

    1976-01-01

    A mathematical simulation of first wall swelling has been performed for stainless steel under a hypothetical duty cycle of 50 sec burn, 50 sec cool. In most instances steady state nucleation conditions were not established during the burn cycle, thereby necessitating the use of transient nucleation theory. The effects of transmutation helium and of surface active impurities were modelled in an approximate way. Both kinds of impurity were found to give large increases in the void nucleation rate. Suggestions for refining and extending the calculations are also given

  16. Limiter and first wall of the fusion reactor blanket

    International Nuclear Information System (INIS)

    Danilov, I.; Skladnov, K.; Kolganov, V.

    1994-01-01

    Previous designing of the first wall and limiter has allowed to determine their possible embodiment depending on the parameters and operation conditions of the blanket. As a rule limiter is a separate structure located on the plasma facing surface of the blanket assembly. Possible versions of the limiter/FW which may be considered: (1) limiters with mechanical attachment of the protective part; (2) limiters with the attachment with brazing; (3) limiters with common/separate cooling system; (4) limiter as a substitute of the FW. Generally the FW/limiter structure includes protective shield and its cooling system which consist of protective coating, heat accumulator, conductive layer and attachment locks

  17. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Huang, J.S.; Guinan, M.W.; Hahn, P.A.

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  19. Fusion neutron irradiation of Ni(Si) alloys at high temperature

    International Nuclear Information System (INIS)

    Huang, J.S.; Guinan, M.W.; Hahn, P.A.

    1987-09-01

    Two Ni-4% Si alloys, with different cold work levels, are irradiated with 14 MeV fusion neutrons at 623 K, and their Curie temperatures are monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2 MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14 MeV fusion neutrons is only 6 to 7% of that for an identical alloy irradiated by 2 MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6 to 7% for the fusion neutron irradiated sample. 17 refs., 4 figs., 1 tab

  20. Fusion neutron irradiation of Ni-Si alloys at high temperature*1

    Science.gov (United States)

    Huang, J. S.; Guinan, M. W.; Hahn, P. A.

    1988-07-01

    Two Ni-4% Si alloys, with different cold work levels, have been irradiated with 14-MeV fusion neutrons at 623 K, and their Curie temperatures have been monitored during irradiation. The results are compared to those of an identical alloy irradiated by 2-MeV electrons. The results show that increasing dislocation density increases the Curie temperature change rate. At the same damage rate, the Curie temperature change rate for the alloy irradiated by 14-MeV fusion neutrons is only 6-7% of that for an identical alloy irradiated by 2-MeV electrons. It is well known that the migration of radiation induced defects contributes to segregation of silicon atoms at sinks in this alloy, causing the Curie temperature changes. The current results imply that the relative free defect production efficiency decreases from one for the electron irradiated sample to 6-7% for the fusion neutron irradiated sample.

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

    Directory of Open Access Journals (Sweden)

    Guler Nevzat

    2013-11-01

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

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

    Science.gov (United States)

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

    2013-11-01

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

  3. Fusion reaction around the Coulomb barrier with neutron-rich nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Atsushi [Institute of Physical and Chemical Research, Wako, Saitama (Japan)

    1997-07-01

    Two fusion reactions with neutron-rich nuclei are reported in this work. On the first reaction: {sup 9,10,11}Be+{sup 209}Bi, the fusion cross sections around the coulomb barrier were measured by determing {alpha} disintegration from compound nucleus Fr. In the field of 10-100 mb, the same total fusion cross sections were obtained. The phenomenon {sup 11}Be(neutron halo nucleus) alone increased and decreased was not observed. The fusion cross sections of {sup 27,29,31}Al+{sup 197}Au system were determined by using 130 kcps and 30 kcps of beam strength of {sup 29,31}Al, respectively. The value of {sup 27}Al was reproduced by calculation, but that of {sup 29}Al increased around barrier which could not be explained by CCDEF calculation. (S.Y.)

  4. Chamber wall response to target implosion in inertial fusion reactors: new and critical assessments

    International Nuclear Information System (INIS)

    Hassanein, A.; Morozov, V.

    2002-01-01

    The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to next target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The focus of this study is to critically assess the reliability and the dynamic response of chamber walls in IFE systems. Of particular concern is the effect on wall erosion lifetime due to various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls

  5. Chamber wall response to target implosion in inertial fusion reactors : new and critical assessments

    International Nuclear Information System (INIS)

    Hassanein, A.; Morozov, V.

    2002-01-01

    The chamber walls in inertial fusion energy (IFE) reactors are exposed to harsh conditions following each target implosion. Key issues of the cyclic IFE operation include intense photon and ion deposition, wall thermal and hydrodynamic evolution, wall erosion and fatigue lifetime, and chamber clearing and evacuation to ensure desirable conditions prior to target implosion. Several methods for wall protection have been proposed in the past, each having its own advantages and disadvantages. These methods include use of solid bare walls, gas-filled cavities, and liquid walls/jets. Detailed models have been developed for reflected laser light, emitted photons, and target debris deposition and interaction with chamber components and have been implemented in the comprehensive HEIGHTS software package. The hydrodynamic response of gas filled cavities and photon radiation transport of the deposited energy has been calculated by means of new and advanced numerical techniques. Fragmentation models of liquid jets as a result of the deposited energy have also been developed, and the impact on chamber clearing dynamics has been evaluated. Th focus of this study is to critically assess the reliability and the dynamic response of chamber walls in various proposed protection methods for IFE systems. Of particular concern is the effect on wall erosion lifetime of various erosion mechanisms, such as vaporization, chemical and physical sputtering, melt/liquid splashing and explosive erosion, and fragmentation of liquid walls

  6. High energy neutron dosimetry for the fusion program

    International Nuclear Information System (INIS)

    Barr, D.W.; Norris, A.E.

    1977-01-01

    Neutron dosimetry by the foil activation method offers a flexible technique for characterizing neutron spectra ranging from thermal energies to 30 MeV with the potential for extension to higher neutron energies as investigated by the Los Alamos Radiochemistry Group at the Los Alamos Meson Physics Facility and in the Apollo-Soyuz Test Project. The use of this method for the neutron flux description in thermal, resonance, and fission spectrum assemblies has been demonstrated. An extension of the method to environments involving thermonuclear processes was developed at Los Alamos in the early 1950's to characterize mixed fission-thermonuclear systems

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

    International Nuclear Information System (INIS)

    Sugimoto, Masayoshi

    2001-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  9. Nuclear science experiments with a bright neutron source from fusion reactions on the OMEGA Laser System

    Science.gov (United States)

    Forrest, C. J.; Knauer, J. P.; Schroeder, W. U.; Glebov, V. Yu.; Radha, P. B.; Regan, S. P.; Sangster, T. C.; Sickles, M.; Stoeckl, C.; Szczepanski, J.

    2018-04-01

    Subnanosecond impulses of 1013 to 1014 neutrons, produced in direct-drive laser inertial confinement fusion implosions, have been used to irradiate deuterated targets at the OMEGA Laser System (Boehly et al., 1997). The target compounds include heavy water (D2O) and deuterated benzene (C6D6). Yields and energy spectra of neutrons from D(n,2n)p to study the breakup reaction have been measured at a forward angle of θlab = 3 .5∘ ± 3.5° with a sensitive, high-dynamic-range neutron time-of-flight spectrometer to infer the double-differential breakup cross section d2 σ/dE d Ω for 14-MeV D-T fusion neutrons.

  10. A Fusion Neutron Source for Materials and Subcomponent Development and Qualification

    Science.gov (United States)

    Simonen, Thomas

    2010-11-01

    The magnetic-mirror based Gas Dynamic Trap (GDT) device in Novosibirsk Russia is developing the physics basis for a compact DT Neutron Source (DTNS) for fusion materials and subcomponent development as well as a driver for a fusion-fission driver for nuclear waste burn-up. The efficiency of this concept depends on electron temperature. This paper describes past experimental results as well as methods and prospects to further increase the electron temperature.

  11. The mechanical performance of the fusion reactor first wall. Pt. 2

    International Nuclear Information System (INIS)

    Daenner, W.; Raeder, J.

    1977-03-01

    While the first part of this report was concerned with the steady-state mechanical analysis of the fusion reactor first wall, this part deals with the analysis based upon pulsed load conditions. In a first section we elaborate various solutions of the non-stationary heat conduction problem in plane geometry capable of describing the temperature response of the wall due to characteristic plasma pulse sequences. these solutions are input to a quasi-steady-state stress and strain analysis. Finally, the results of this analysis are set in relation to the fatigue properties of the wall material. A further section presents a description of a computer program which uses the mathematical procedure described. The results of some test runs are followed by those of detailed parameter studies. In the course of these calculations the influences of a number of design and operational quantities of a fusion reactor were investigated. It turned out that the choice of wall thickness and wall loading are of predominant importance for the first wall fatigue life. (orig.) [de

  12. Interactions of D-T neutrons in graphite and lithium blankets of fusion reactors

    International Nuclear Information System (INIS)

    Ofek, R.

    1986-05-01

    The present study deals with integral experiment and calculation of neutron energy spectra in bulks of graphite which is used as a reflector in blankets of fusion reactors, and lithium, the material of the blanket on which lithium is bred due to neutron interactions. The collimated beam configuration enables - due to the almost monoenergeticity and unidirectionality of the neutrons impinging on the target - to identify fine details in the measured spectra, and also facilitates the absolute normalization of the spectra. The measured and calculated spectra are generally in a good agreement and in a very good agreement at mesh points close to the system axis. A few conclusions may be drawn: a) the collimated beam source configuration is a sensitive tool for measuring neutron energy spectra with a high resolution, b) the method of unfolding proton-recoil spectra measured with a NE-213 scintillator should be improved, c) MCNP and DOT 4.2 may be used as complementary codes for neutron transport calculations of fusion blankets and deep-penetration problems, d) the updating of the cross-section libraries and checking by integral experiments is highly important for the design of fusion blankets. The present study may be regarded as an important course in the research and development of tools for the design of fusion blankets

  13. Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

    International Nuclear Information System (INIS)

    Ogawa, Y; Goto, T; Okano, K; Asaoka, Y; Hiwatari, R; Someya, Y

    2008-01-01

    The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G∼100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ∼ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive

  14. Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Y [High Temperature Plasma Center, University of Tokyo, Chiba (Japan); Goto, T; Okano, K [Graduate School of Frontier Sciences, University of Tokyo, Chiba (Japan); Asaoka, Y; Hiwatari, R [Central Research Institute for Electric Power Industry, Komae, Tokyo (Japan); Someya, Y [Graduate School of Engineering, Musashi Institute of Technology, Tokyo (Japan)], E-mail: ogawa@ppl.k.u-tokyo.ac.jp

    2008-05-15

    The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G{approx}100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 {approx} 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

  15. Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

    Science.gov (United States)

    Ogawa, Y.; Goto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Someya, Y.

    2008-05-01

    The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G~100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ~ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

  16. High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

    1996-10-01

    A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

  17. Fusion Power Measurement Using a Combined Neutron Spectrometer-Camera System at ITER

    International Nuclear Information System (INIS)

    Sjoestrand, Henrik; Sunden, E. Andersson; Conroy, S.; Ericsson, G.; Johnson, M. Gatu; Giacomelli, L.; Hellesen, C.; Hjalmarsson, A.; Ronchi, E.; Weiszflog, M.; Kaellne, J.

    2008-01-01

    A central task for fusion plasma diagnostics is to measure the 2.5 and 14 MeV neutron emission rate in order to determine the fusion power. A new method for determining the neutron yield has been developed at JET. It makes use of the magnetic proton recoil neutron spectrometer and a neutron camera and provides the neutron yield with small systematic errors. At ITER a similar system could operate if a high-resolution, high-performance neutron spectrometer similar to the MPR was installed. In this paper, we present how such system could be implemented and how well it would perform under different assumption of plasma scenarios and diagnostic capabilities. It is found that the systematic uncertainty for using such a system as an absolute calibration reference is as low as 3% and hence it would be an excellent candidate for the calibration of neutron monitors such as fission chambers. It is also shown that the system could provide a 1 ms time resolved estimation of the neutron rate with a total uncertainty of 5%

  18. Collimator design for neutron imaging of laser-fusion targets

    International Nuclear Information System (INIS)

    Sommargren, G.E.; Lerche, R.A.

    1981-01-01

    Several pinhole collimator geometries for use in neutron imaging experiments have been modeled and compared. Point spread functions are shown for a cylinder, hyperbola, intersecting cones, and a five-zone approximation to the intersecting cones. Of the geometries studied, the intersecting cones appear the most promising with respect to neutron efficiency, field of view, and isoplanatism

  19. Neutron spectroscopy on TFTR [Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Nishitani, T.; Strachan, J.D.

    1988-05-01

    This paper describes the use of an 3 He ionization chamber for neutron spectroscopy on TFTR during 1987. The ion temperature was measured using neutron spectroscopy for one set of ohmically heated plasmas. The deduced ion temperatures agreed to within 20% with those measured by other diagnostics. 11 refs., 11 figs., 1 tab

  20. Numerical study of neutron beam divergence in a beam-fusion scenario employing laser driven ions

    Science.gov (United States)

    Alejo, A.; Green, A.; Ahmed, H.; Robinson, A. P. L.; Cerchez, M.; Clarke, R.; Doria, D.; Dorkings, S.; Fernandez, J.; McKenna, P.; Mirfayzi, S. R.; Naughton, K.; Neely, D.; Norreys, P.; Peth, C.; Powell, H.; Ruiz, J. A.; Swain, J.; Willi, O.; Borghesi, M.; Kar, S.

    2016-09-01

    The most established route to create a laser-based neutron source is by employing laser accelerated, low atomic-number ions in fusion reactions. In addition to the high reaction cross-sections at moderate energies of the projectile ions, the anisotropy in neutron emission is another important feature of beam-fusion reactions. Using a simple numerical model based on neutron generation in a pitcher-catcher scenario, anisotropy in neutron emission was studied for the deuterium-deuterium fusion reaction. Simulation results are consistent with the narrow-divergence (∼ 70 ° full width at half maximum) neutron beam recently served in an experiment employing multi-MeV deuteron beams of narrow divergence (up to 30° FWHM, depending on the ion energy) accelerated by a sub-petawatt laser pulse from thin deuterated plastic foils via the Target Normal Sheath Acceleration mechanism. By varying the input ion beam parameters, simulations show that a further improvement in the neutron beam directionality (i.e. reduction in the beam divergence) can be obtained by increasing the projectile ion beam temperature and cut-off energy, as expected from interactions employing higher power lasers at upcoming facilities.

  1. Activation Inventories after Exposure to DD/DT Neutrons in Safety Analysis of Nuclear Fusion Installations.

    Science.gov (United States)

    Stankunas, Gediminas; Cufar, Aljaz; Tidikas, Andrius; Batistoni, Paola

    2017-11-23

    Irradiations with 14 MeV fusion neutrons are planned at Joint European Torus (JET) in DT operations with the objective to validate the calculation of the activation of structural materials in functional materials expected in ITER and fusion plants. This study describes the activation and dose rate calculations performed for materials irradiated throughout the DT plasma operation during which the samples of real fusion materials are exposed to 14 MeV neutrons inside the JET vacuum vessel. Preparatory activities are in progress during the current DD operations with dosimetry foils to measure the local neutron fluence and spectrum at the sample irradiation position. The materials included those used in the manufacturing of the main in-vessel components, such as ITER-grade W, Be, CuCrZr, 316 L(N) and the functional materials used in diagnostics and heating systems. The neutron-induced activities and dose rates at shutdown were calculated by the FISPACT code, using the neutron fluxes and spectra that were provided by the preceding MCNP neutron transport calculations. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  2. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2014-01-01

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM

  3. Physical Investigation for Neutron Consumption and Multiplication in Blanket Module of Fusion-Fission Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tariq Siddique, M.; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-05-15

    Fusion-fission hybrid reactor can be the first milestone of fusion technology and achievable in near future. It can provide operational experience for tritium recycling for pure fusion reactor and be used for incineration of high-level long-lived waste isotopes from existing fission power reactors. Hybrid reactor for waste transmutation (Hyb-WT) was designed and optimized to assess its otential for waste transmutation. ITER will be the first large scaled experimental tokamak facility for the testing of test blanket modules (TBM) which will layout the foundation for DEMO fusion power plants. Similarly hybrid test blanket module (HTBM) will be the foundation for rationality of fusion fission hybrid reactors. Designing and testing of hybrid blankets will lead to another prospect of nuclear technology. This study is initiated with a preliminary design concept of a hybrid test blanket module (HTBM) which would be tested in ITER. The neutrons generated in D-T fusion plasma are of high energy, 14.1 MeV which could be multiplied significantly through inelastic scattering along with fission in HTBM. In current study the detailed neutronic analysis is performed for the blanket module which involves the neutron growth and loss distribution within blanket module with the choice of different fuel and coolant materials. TRU transmutation and tritium breeding performance of HTBM is analyzed under ITER irradiation environment for five different fuel types and with Li and LiPb coolants. Simple box geometry with plate type TRU fuel is adopted so that it can be modelled with heterogeneous material geometry in MCNPX. Waste transmutation ratio (WTR) of TRUs and tritium breeding ration (TBR) is computed to quantify the HTBM performance. Neutron balance is computed in detail to analyze the performance parameters of HTBM. Neutron spectrum and fission to capture ratio in TRU fuel types is also calculated for detailed analysis of HTBM.

  4. Change in properties of superconducting magnet materials by fusion neutron irradiation

    International Nuclear Information System (INIS)

    Nishimura, Arata; Nishijima, Shigehiro; Takeuchi, Takao; Nishitani, Takeo

    2007-01-01

    A fusion reactor will generate a lot of high energy neutron and much energy will be taken out of the neutrons by a blanket system. Since some neutrons will stream out of a plasma vacuum vessel through neutral beam injection ports and penetrate a blanket system, a superconducting magnet system, which provides high magnetic field to confirm high energy particles, will be irradiated by a certain amount of neutrons. By developing the new NBI system or by reducing the penetration, the neutron fluence to the superconducting magnet will be able to be reduced. However, it is not easy to achieve the lower streaming and penetration at the present. Therefore, investigations on irradiation behavior of superconducting magnet materials are desired and some novel researches have been performed from 1970s. In general, the critical current of the superconducting wire increases under fast neutron environment comparing with that of the non-irradiated wire, and then decreased to almost zero as an increase of neutron fluence. On the other hand, the critical temperature of the wire starts to get down around 10 22 n/m 2 of neutron fluence and the temperature margin will be decreased during the operation by the neutron irradiation. In this paper, some aspects of irradiated materials will be overviewed and general tendency will be discussed focussing on knock-on effect of fast neutron and long range ordering of A15 compounds

  5. Neutron irradiation experiments for fusion reactor materials through JUPITER program

    International Nuclear Information System (INIS)

    Abe, K.; Namba, C.; Wiffen, F.W.; Jones, R.H.

    1998-01-01

    A Japan-USA program of irradiation experiments for fusion research, ''JUPITER'', has been established as a 6 year program from 1995 to 2000. The goal is to study ''the dynamic behavior of fusion reactor materials and their response to variable and complex irradiation environment''. This is phase-three of the collaborative program, which follows RTNS-II program (phase-1: 1982-1986) and FFTF/MOTA program (phase-2: 1987-1994). This program is to provide a scientific basis for application of materials performance data, generated by fission reactor experiments, to anticipated fusion environments. Following the systematic study on cumulative irradiation effects, done through FFTF/MOTA program. JUPITER is emphasizing the importance of dynamic irradiation effects on materials performance in fusion systems. The irradiation experiments in this program include low activation structural materials, functional ceramics and other innovative materials. The experimental data are analyzed by theoretical modeling and computer simulation to integrate the above effects. (orig.)

  6. Detail analysis of fusion neutronics benchmark experiment on beryllium

    International Nuclear Information System (INIS)

    Konno, Chikara; Ochiai, Kentaro; Takakura, Kosuke; Ohnishi, Seiki; Kondo, Keitaro; Wada, Masayuki; Sato, Satoshi

    2010-01-01

    Our previous analysis of the integral experiments (in situ and TOF experiments) on beryllium with DT neutrons at JAEA/FNS pointed out two problems by using MCNP4C and the latest nuclear data libraries; one was a strange larger neutron peak around 12 MeV appearing in the TOF experiment analysis with JEFF-3.1 and the other was an overestimation on law energy neutrons in the in situ experiment analyses with all the nuclear data libraries. We investigated reasons for these problems in detail. It was found out that the official ACE file MCJEFF3.1 of JEFF-3.1 had an inconsistency with the original JEFF-3.1, which caused the strange larger neutron peak around 12 MeV in the TOF experiment analysis. We also found out that the calculated thermal neutron peak was probably too large in the in situ experiment. On trial we examined influence of the thermal neutron scattering law data of beryllium metal in ENDF/B-VI. The result pointed out that the coherent elastic scattering cross-section data in the thermal neutron scattering law data of beryllium metal were probably too large.

  7. Studies of neutron measurement methods for fusion plasma diagnostics

    International Nuclear Information System (INIS)

    Beimer, K.H.

    1986-03-01

    This thesis comprises several studies mainly devoted to neutron measurement systems for plasma diagnostics at JET (Joint European Torus). An in situ calibration of the U-235 fission chamber detectors located at JET is presented. These detectors are used for measuring the neutron yield from the thermonuclear reactions in the plasma. The energy spectrum of the neutrons from the reactions D(d,n) 3 He has been studied by means of a 3 He spectrometer. Especially, it was found that by measuring the width of the full energy peak in the response spectrum of the 3 He-spectrometer, the deuterium distribution in the deuterium targets used can be estimated. In order to measure different neutron energies it is necessary to obtain a detailed knowledge of the response of the spectrometer. Therefore, the response function to monoenergetic neutrons in the energy range 130-3030 keV was experimentally determined. Some work has been related to a design study of a 14 MeV spectrometer for neutron diagnostics. It is a combined proton-recoil and time-of-flight spectrometer for high resolution measurements. The main parts of it are the collimator, the scattering foil, and the detectors for the recoil protons and the scattered neutrons. The influence of proton straggling in the foil on the resolution and efficiency of the spectrometer has been studied. Furthermore, a three dimensional Monte Carlo code has been written and used for the design of the collimator. (author)

  8. Backtracing neutron analysis in the fusion-fission dynamics study

    International Nuclear Information System (INIS)

    Brennand, E. de Goes; Hanappe, F.; Stuttge, L.

    2001-01-01

    A new method for the analysis of multi parametric experimental data is used in the study of the dynamics of the fission process for the compound system 126 Ba. We apply this method to obtain the correlation between thermal energy related to the neutron total multiplicity and the correlation between pre-scission neutron and pos-scission neutron multiplicities. The results obtained are interpreted into the framework of a dynamical model. From this interpretation we have access to the following information: the friction intensity which drives the dynamical evolution of the system; the initial deformation of the compound system; the barrier evolution with temperature and angular momentum, and fission times. (author)

  9. Benchmarking of the FENDL-3 Neutron Cross-section Data Starter Library for Fusion Applications

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, U., E-mail: ulrich.fischer@kit.edu [Association KIT-Euratom, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Angelone, M. [Associazione ENEA-Euratom, ENEA Fusion Division, Via E. Fermi 27, I-00044 Frascati (Italy); Bohm, T. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States); Kondo, K. [Association KIT-Euratom, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Konno, C. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195 (Japan); Sawan, M. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States); Villari, R. [Associazione ENEA-Euratom, ENEA Fusion Division, Via E. Fermi 27, I-00044 Frascati (Italy); Walker, B. [University of Wisconsin-Madison, 1500 Engineering Dr, Madison, WI 53706 (United States)

    2014-06-15

    This paper summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) on a computational ITER benchmark and a series of 14 MeV neutron benchmark experiments. The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses. In general, FENDL-3 shows an improved performance for fusion neutronics applications.

  10. NSPEC - A neutron spectrum code for beam-heated fusion plasmas

    International Nuclear Information System (INIS)

    Scheffel, J.

    1983-06-01

    A 3-dimensional computer code is described, which computes neutron spectra due to beam heating of fusion plasmas. Three types of interactions are considered; thermonuclear of plasma-plasma, beam-plasma and beam-beam interactions. Beam deposition is modelled by the NFREYA code. The applied steady state beam distribution as a function of pitch angle and velocity contains the effects of energy diffusion, friction, angular scattering, charge exchange, electric field and source pitch angle distribution. The neutron spectra, generated by Monte-Carlo methods, are computed with respect to given lines of sight. This enables the code to be used for neutron diagnostics. (author)

  11. A study of hydrogen isotopes fuel control by wall effect in magnetic fusion devices

    Energy Technology Data Exchange (ETDEWEB)

    Motevalli, S.M., E-mail: motavali@umz.ac.ir; Safari, M.

    2016-11-15

    Highlights: • A particle balance model for the main plasma and wall inventory in magnetic fusion device has been represented. • The dependence of incident particles energy on the wall has been considered in 10–300 eV for the sputtering yield and recycling coefficient. • The effect of fueling methods on plasma density behavior has been studied. - Abstract: Determination of plasma density behavior in magnetic confinement system needs to study the plasma materials interaction in the facing components such as first wall, limiter and divertor. Recycling of hydrogen isotope is an effective parameter in plasma density rate and plasma fueling. Recycling coefficient over the long pulse operation, gets to the unity, so it has a significant effect on steady state in magnetic fusion devices. Typically, sputtered carbon atoms from the plasma facing components form hydrocarbons and they redeposit on the wall. In this case little rate of hydrogen loss occurs. In present work a zero dimensional particle equilibrium model has been represented to determine particles density rate in main plasma and wall inventory under recycling effect and codeposition of hydrogen in case of continues and discontinues fueling methods and effective parameters on the main plasma decay has been studied.

  12. Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-02-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate.

  13. Damage of first wall materials in fusion reactors under nonstationary thermal effects

    International Nuclear Information System (INIS)

    Maslaev, S.A.; Platonov, Yu.M.; Pimenov, V.N.

    1991-01-01

    The temperature distribution in the first wall of a fusion reactor was calculated for nonstationary thermal effects of the type of plasma destruction or the flow of 'running electrons' taking into account the melting of the surface layer of the material. The thickness of the resultant damaged layer in which thermal stresses were higher than the tensile strength of the material is estimated. The results were obtained for corrosion-resisting steel, aluminium and vanadium. Flowing down of the molten layer of the material of the first wall is calculated. (author)

  14. Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

    International Nuclear Information System (INIS)

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

    1981-02-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate

  15. Mechanical and microstructural characterization of low activation steels as first wall of nuclear fusion reactors

    International Nuclear Information System (INIS)

    Hernandez, M.T.; Lapena, J.; Diego, G. de; Schirra, M.

    1996-01-01

    Currently, the design development of fusion reactors and the possible materials to use in them are being studied in parallel. One of the most critical problems in this research is the structural materials selection for the first wall and blanket. The aim of the present work is to study three low activation alloys designed in Germany in which niobium has been substituted by tantalum or cerium. The mechanical results show that the alloys containing cerium are in the same order of the low activation materials known to date, but the tantalum doped alloy produces TaC 3 precipitation that destabilizes the matrix and provokes large microstructural changes. This causes a decrease of the mechanical properties at about 600 degree centigree. This fact makes this alloy insuitable for the first wall on fusion reactors, because the working temperature is near 550 degree centigree. (Author) 11 refs

  16. Neutron energy spectrum in graphite blankets of fusion reactors

    International Nuclear Information System (INIS)

    Tsechanski, A.

    1981-09-01

    Neutron flux measurements were performed in a graphite stack and compared with calculations made with a two dimensional transport computer code. In the present work it is observed that the calculated spectrum in the elastic and inelastic scattering ranges (the first collision range in both cases), is sensitive to details of the angular distribution of these neutrons. Regarding the discrepancies in the elastic scattering range it is concluded that the microscopic cross section library ENDF/B-IV overestimates the large angle scattering (back scattering) as can be seen from comparison of measured and calculated spectra. The two most important conclusions of the present work are: 1. Inelastic scattering interaction of D-T neutrons in graphite cannot be calculated without a proper account of energy-angle correlation. 2. An experimental setup supplying monoenergetic collimated D-T neutrons constitutes a sensitive although indirect means for measuring angular distributions in inelastic and elastic scattering

  17. DT fusion neutron irradiation of LLL Nb3Sn and LLL superconductor wires at 4.20K

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation of one LLL superconductor wire and one LLL Nb 3 Sn foil at 4.2 0 K is described. The sample position, beam-on time, and neutron dose record are given. The results from two ''profile'' dosimetry foils measuring the lateral variation in neutron flux are included

  18. Evaluation of neutron streaming through injection ports in a tokamak-type fusion reactor

    International Nuclear Information System (INIS)

    Ide, Takahiro; Seki, Yasushi; Iida, Hiromasa

    1976-03-01

    The effects of neutron streaming through injection ports in the fusion reactor designed in JAERI have been studied, especially those on tritium breeding ratio and the shielding of the superconducting magnet. In placement of the injection ports in the blanket, the tritium breeding ratio decreases by up to 1.3%, and shielding problem of the superconducting magnet is very important. (auth.)

  19. Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao

    1978-01-01

    An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)

  20. Heat transfer phenomena in the first wall of the RFX fusion experiment

    International Nuclear Information System (INIS)

    Oliveira Pauletti, R.M. de

    1988-12-01

    The thermal analysis of the first wall (FW) of the RFX machine is presented. RFX is a large fusion experiment under construction at Padua, Italy. The RFX FW is briefly described, together with the critical thermal conditions. The numerical analyses performed to predict the FW thermal behaviour are presented. 1-D and 2-D finite element models give accurate predictions of the FW temperatures and of the thermal exchanges in the machine environment. (author) [pt

  1. Stagnation of ablated metal vapor in laser fusion reactor with liquid wall

    International Nuclear Information System (INIS)

    Norimatsu, T.; Nagatomo, H.; Azechi, H.; Furukawa, H.; Shimada, Y.; Kurahashi, S.; Kunugi, T.; Kajimura, Y.

    2010-11-01

    In this paper, formation of clusters by ablated materials and those stagnation at the center of a laser fusion reactor with liquid wall are discussed using improved simulation code DECORE. We will report 1) numerical simulation on formation of clusters immediately before the stagnation, 2) preliminary results on the cluster formation at the first bounce of the stagnation, 3) experimental result on the diameter measurement of micro droplets formed in a simulation experiment with back-side irradiation of laser. (author)

  2. The development of a spectrometer for 14 MeV neutrons from fusion

    International Nuclear Information System (INIS)

    Aronsson, D.

    1991-01-01

    A spectrometer for 14 MeV neutrons, to be used for fusion plasma diagnostics at JET, was developed. The spectrometer utilizes neutron scattering in a polyethylene foil with the detection of the scattered neutron and its associated recoil proton. For the detection of 12 MeV protons we have tested silicon surface barrier detectors, lithium-drifted silicon detectors and high purity germanium detectors. The lithium-drifted detectors were finally selected for use in the spectrometer. The lithium-drifted silicon diodes have also been used for direct spectrometry, utilizing the neutron induced charged particle reactions in silicon. The methods used for the energy calibration and the timing calibration of the diodes, both during the installation of the spectrometer and during operation, are described. The detection of 2 MeV neutrons is done by fast plastic scintillators. Since the neutron generator which was used to test the detectors supplies 14 MeV or 2.5 MeV neutrons only, a neutron energy converter has to be constructed to study the detectors at other neutron energies. In the actual spectrometer an array of scintillation neutron detectors is used. A method of calibrating such an array of detectors with a gamma source was elaborated and is also described here. The result of the calibration is a set of parameters than can be used to determine the high voltage settings and the discriminator levels that are needed to achieve homogeneous sensitivity for all the detectors of the array. The energy scale itself was then calibrated by using gamma sources of various energies. To test the spectrometer as a whole at a neutron generator, a test bed was constructed. A lithium-drifted silicon diode was used to measure the neutron flux and the neutron energy resolution in the test bed. (au)

  3. Development of High Intensity D-T fusion NEutron Generator (HINEG)

    Science.gov (United States)

    Wu, Yican; Liu, Chao; Song, Gang; Wang, Yongfeng; Li, Taosheng; Jiang, Jieqiong; Song, Yong; Ji, Xiang

    2017-09-01

    A high intensity D-T fusion neutron generator (HINEG) is keenly needed for the research and development (R&D) of nuclear technology and safety of the advanced nuclear energy system, especially for the radiation protection and shielding. The R&D of HINEG includes two phases: HINEG-I and HINEG-II. HINEG-I is designed to have both the steady beam and pulsed beam. The neutron yield of the steady beam is up to 1012 n/s. The width of pulse neutron beam is less than 1.5 ns. HINEG-I is used for the basic neutronics study, such as measurement of nuclear data, validation of neutronics methods and software, validation of radiation protection and so on. HINEG-II aims to generate a high neutron yield of 1013 n/s neutrons by adopting high speed rotating tritium target system integrated with jet/spray array enhanced cooling techniques, and can further upgrade to obtain neutron yield of 1014 1015n/s by using of accelerators-array in a later stage. HINEG-II can be used for fundamentals research of nuclear technology including mechanism of materials radiation damage and neutronics performance of components, radiation shielding as well as other nuclear technology applications.

  4. Development of High Intensity D-T fusion NEutron Generator (HINEG

    Directory of Open Access Journals (Sweden)

    Wu Yican

    2017-01-01

    Full Text Available A high intensity D-T fusion neutron generator (HINEG is keenly needed for the research and development (R&D of nuclear technology and safety of the advanced nuclear energy system, especially for the radiation protection and shielding. The R&D of HINEG includes two phases: HINEG-I and HINEG-II. HINEG-I is designed to have both the steady beam and pulsed beam. The neutron yield of the steady beam is up to 1012 n/s. The width of pulse neutron beam is less than 1.5 ns. HINEG-I is used for the basic neutronics study, such as measurement of nuclear data, validation of neutronics methods and software, validation of radiation protection and so on. HINEG-II aims to generate a high neutron yield of 1013 n/s neutrons by adopting high speed rotating tritium target system integrated with jet/spray array enhanced cooling techniques, and can further upgrade to obtain neutron yield of 1014~1015n/s by using of accelerators-array in a later stage. HINEG-II can be used for fundamentals research of nuclear technology including mechanism of materials radiation damage and neutronics performance of components, radiation shielding as well as other nuclear technology applications.

  5. Neutron-induced reactions relevant for Inertial-Cofinement Fusion Experiments

    Science.gov (United States)

    Boswell, Melissa; Devlin, Mathew; Fotiadis, Nikolaos; Merrill, Frank; Nelson, Ronald; Tonchev, Anton

    2014-09-01

    The typical ignition experiment at the National Ignition Facility ablatively implodes a plastic capsule filled with DT fuel, generating a high flux of 14-MeV neutrons from the d(t,n) α reaction. There is some spread in the energy of these primary 14-MeV neutrons, which is mainly attributable to Doppler shifting from the relative thermal motion of the burning DT fuel. Neutrons created during this reaction have 5--10% chance of scattering before escaping the fuel assembly, losing some fraction of their energy in the scattering process. Neutrons emerging with an energy greater than the reaction energy are generated by a two-step process where neutrons first transfer momentum to a deuteron or tritium ion, these enhanced energy ions then fuse in flight to produce higher energy neutrons; some of these neutrons have energies in excess of 30 MeV. Measuring the fluencies of both the low- and high-energy neutrons is a powerful mechanism for studying the properties of the fuel assembly, and the various parameters important to inertial confinement fusion. We have developed a number of tools to measure the spectral characteristics of the NIF neutron spectrum. Most of these methods rely on exploiting the energy dependence of (n, γ), (n,2n), (n,3n) and (n,p) reactions on a variety o.

  6. Radiation Shielding Information Center: a source of computer codes and data for fusion neutronics studies

    International Nuclear Information System (INIS)

    McGill, B.L.; Roussin, R.W.; Trubey, D.K.; Maskewitz, B.F.

    1980-01-01

    The Radiation Shielding Information Center (RSIC), established in 1962 to collect, package, analyze, and disseminate information, computer codes, and data in the area of radiation transport related to fission, is now being utilized to support fusion neutronics technology. The major activities include: (1) answering technical inquiries on radiation transport problems, (2) collecting, packaging, testing, and disseminating computing technology and data libraries, and (3) reviewing literature and operating a computer-based information retrieval system containing material pertinent to radiation transport analysis. The computer codes emphasize methods for solving the Boltzmann equation such as the discrete ordinates and Monte Carlo techniques, both of which are widely used in fusion neutronics. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

  7. Research and Development of Landmine Detection System by a Compact Fusion Neutron Source

    International Nuclear Information System (INIS)

    Yoshikawa, Kiyoshi; Masuda, Kai; Toku, Hisayuki; Nagasaki, Kazunobu; Mizutani, Toshiyuki; Takamatsu, Teruhisa; Imoto, Masaki; Yamamoto, Yasushi; Ohnishi, Masami; Osawa, Hodaka; Hotta, Eiki; Kohno, Toshiyuki; Okino, Akitoshi; Watanabe, Masato; Yamauchi, Kunihito; Yuura, Morimasa; Shiroya, Seiji; Misawa, Tsuyoshi; Mori, Takamasa

    2005-01-01

    Current results are described on the research and development of an advanced anti-personnel landmine detection system by using a compact discharge-type fusion neutron source called IECF (Inertial-Electrostatic Confinement Fusion). Landmines are to be identified through backscattering of neutrons, and specific-energy capture γ-rays by hydrogen and nitrogen atoms in the landmine explosives.For this purpose, improvements in the IECF were made by various methods to achieve a drastic enhancement of neutron yields of more than 10 8 n/s in pulsed operation. This required R and D on the power source, as well as analysis of envisaged detection systems with multi-sensors. The results suggest promising and practical features for humanitarian landmine detection, particularly, in Afghanistan

  8. Inertial electrostatic confinement fusion neutron source R ampersand D and issues

    International Nuclear Information System (INIS)

    Ohnishi, Masami; Yamamoto, Yasushi; Hasegawa, Mitsunori

    1997-01-01

    An inertial electrostatic confinement (IEC) fusion is the scheme of injecting the ions and electrons toward the spherical center, trapping both species in the electrostatic self-field and giving rise to fusion reactions in the dense core. An IEC is expected to have wide application from a small neutron source to a D- 3 He fusion reactor. Hirsch reported 10 9 n/s deuterium-tritium (D-T) neutron production in the device equipped with ion guns. Recently, Gu et al. measured 10 6 n/s using a D 2 gas discharge between the spherical wire cathode and the anode vacuum vessel, where the applied voltage is 60 kV and the current is 15 mA. We have also obtained similar neutron production at a lower voltage, ∼45 kV in a single-grid IEC device. Fusion reaction rates obtained by IEC experiments so far cannot be explained by a model of a simple potential well structure because the electrical potential peaked at the center prevents making a dense core. Hirsch proposed a multiwell structure called open-quotes poissorsclose quotes to explain the experiments. It is generally believed that there may be some correlation between the potential well structure and the neutron production rate. The scaling of neutron production on the injected ion current is a most important aspect of the problem for the prospect of utilizing IEC for fusion energy. The potential structure and its behavior are keys to the physics in understanding the principle of an IEC

  9. Strong neutron sources - How to cope with weapon material production capabilities of fusion and spallation neutron sources?

    International Nuclear Information System (INIS)

    Englert, M.; Franceschini, G.; Liebert, W.

    2013-01-01

    In this article we investigate the potential and relevance for weapon material production in future fusion power plants and spallation neutron sources (SNS) and sketch what should be done to strengthen these technologies against a non-peaceful use. It is shown that future commercial fusion reactors may have military implications: first, they provide an easy source of tritium for weapons, an element that does not fall under safeguards and for which diversion from a plant could probably not be detected even if some tritium accountancy is implemented. Secondly, large fusion reactors - even if not designed for fissile material breeding - could easily produce several hundred kg Pu per year with high weapon quality and very low source material requirements. If fusion-only reactors will prevail over fission-fusion hybrids in the commercialization phase of fusion technology, the safeguard challenge will be more of a legal than of a technical nature. In pure fusion reactors (and in most SNS) there should be no nuclear material present at any time by design. The presence of undeclared nuclear material would indicate a military use of the plant. This fact offers a clear-cut detection criterion for a covert use of a declared facility. Another important point is that tritium does not fall under the definition of 'nuclear material', so a pure fusion reactor or a SNS that do not use nuclear materials are not directly falling under any international non-proliferation treaty requirements. Non-proliferation treaties have to be amended to take into account that fact. (A.C.)

  10. IFMIF [International Fusion Materials Irradiation Facility], an accelerator-based neutron source for fusion components irradiation testing: Materials testing capabilities

    International Nuclear Information System (INIS)

    Mann, F.M.

    1988-08-01

    The International Fusion Materials Irradiation Facility (IFMIF) is proposed as an advanced accelerator-based neutron source for high-flux irradiation testing of large-sized fusion reactor components. The facility would require only small extensions to existing accelerator and target technology originally developed for the Fusion Materials Irradiation Test (FMIT) facility. At the extended facility, neutrons would be produced by a 0.1-A beam of 35-MeV deuterons incident upon a liquid lithium target. The volume available for high-flux (>10/sup 15/ n/cm/sup 2/-s) testing in IFMITF would be over a liter, a factor of about three larger than in the FMIT facility. This is because the effective beam current of 35-MeV deuterons on target can be increased by a factor of ten to 1A or more. Such an increase can be accomplished by funneling beams of deuterium ions from the radio-frequency quadruple into a linear accelerator and by taking advantage of recent developments in accelerator technology. Multiple beams and large total current allow great variety in available testing. For example, multiple simultaneous experiments, and great flexibility in tailoring spatial distributions of flux and spectra can be achieved. 5 refs., 2 figs., 1 tab

  11. Pycnonuclear fusion rates. [In white dwarf and neutron stars

    Energy Technology Data Exchange (ETDEWEB)

    Schramm, S.; Koonin, S.E. (California Institute of Technology, Pasadena (USA))

    1990-12-01

    Pycnonuclear fusion processes in an ionic crystal at zero temperature are discussed. Rearrangement effects of the lattice during the fusion process are accounted for exactly in the screening potential by computing the relaxation of the surrounding nuclei. The dynamics of the rearranging ions also generates an increased effective mass that must be used in describing the tunneling of the fusing nuclei. The opposing effects of lattice polarization through enhanced screening and the effective mass results in pycnonuclear reaction rates that are several orders of magnitude smaller than those calculated in the Wigner-Seitz approximation and are roughly equal to those obtained in the static approximation that neglects polarization entirely. 12 refs.

  12. Neutron rich clusters and the dynamics of fission and fusion

    International Nuclear Information System (INIS)

    Armbruster, P.

    1988-07-01

    In this lecture I want to discuss experimental evidence for the appearance of cluster aspects in the dynamics of large rearrangement processes, as fusion and fission. Clusters in the sense as used in my lecture are the strongly bound doubly magic nuclei as 20 Ca 28 48 , 28 Ni 50 78 , 132 50 Sn 82 , and 208 82 Pb 126 and the superheavy nucleus 298 114 184 . Two of these nuclei, 78 Ni and 298 114 have not yet been identified. I discuss first the experimental findings from heavy element production. Then I cover the stability of cluster aspects to intrinsic excitation energy in fusion and fission. (orig./HSI)

  13. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    International Nuclear Information System (INIS)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B.

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si 3 N 4 . Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation

  14. Joining and fabrication techniques for high temperature structures including the first wall in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Lee, B. S.; Kim, K. B

    2003-09-01

    The materials for PFC's (Plasma Facing Components) in a fusion reactor are severely irradiated with fusion products in facing the high temperature plasma during the operation. The refractory materials can be maintained their excellent properties in severe operating condition by lowering surface temperature by bonding them to the high thermal conducting materials of heat sink. Hence, the joining and bonding techniques between dissimilar materials is considered to be important in case of the fusion reactor or nuclear reactor which is operated at high temperature. The first wall in the fusion reactor is heated to approximately 1000 .deg. C and irradiated severely by the plasma. In ITER, beryllium is expected as the primary armour candidate for the PFC's; other candidates including W, Mo, SiC, B4C, C/C and Si{sub 3}N{sub 4}. Since the heat affected zones in the PFC's processed by conventional welding are reported to have embrittlement and degradation in the sever operation condition, both brazing and diffusion bonding are being considered as prime candidates for the joining technique. In this report, both the materials including ceramics and the fabrication techniques including joining technique between dissimilar materials for PFC's are described. The described joining technique between the refractory materials and the dissimilar materials may be applicable for the fusion reactor and Generation-4 future nuclear reactor which are operated at high temperature and high irradiation.

  15. Low-Z coating as a first wall of nuclear fusion devices

    International Nuclear Information System (INIS)

    Shikama, Tatsuo; Okada, Masatoshi

    1984-01-01

    The tokamak nuclear fusion devices of the largest scale in the world, TFTR in USA and JET in Europe, started the operation from the end of 1982 to 1983. Also in Japan, the tokamak JT-60 is scheduled to begin the operation in 1985. One of the technological obstacles is the problem of first walls facing directly to plasma and subjected to high particle loading and thermal loading. Moreover, first walls achieve the active role of controlling impurities in plasma and recycling hydrogen isotopes. It is impossible to find a single material which satisfies all these requirements. The compounding of materials can create a material having new function, but also has the meaning of expanding the range of material selection. One of the material compounding methods is surface coating. In this paper, as the materials for first walls, the characteristics of low Z materials are discussed from the design examples of actual takamak nuclear fusion devices. The outline of first walls is explained. High priority is given to the impurity control in plasma, and in view of plasma energy emissivity and the rate of self sputtering, low Z material coating seems to be the solution. The merits and the problems of such low Z material coating are discussed. (Kako, I.)

  16. Intense neutron source facility for the fusion energy program

    International Nuclear Information System (INIS)

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

    1975-01-01

    The intense neutron source is based on the ability of a supersonic flow of gas to dissipate an enormous quantity of heat generated in the neutron-producing target by multiple Coulomb collisions. A description is given of the principles involved in forming the supersonic jet, in forming the intense tritium-ion beam, in the vacuum systems, and in the tritium handling systems. An overview of the entire facility is included. It is believed that the facility can be operated with high reliability, ensuring a productive radiation damage program. (U.S.)

  17. Measurements of fusion neutron multiplication in spherical beryllium shells

    International Nuclear Information System (INIS)

    Giese, H.; Kappler, F.; Tayama, R.; Moellendorff, U. von; Alevra, A.; Klein, H.

    1996-01-01

    New results of spherical-shell transmission measurements with 14-MeV neutrons on pure beryllium shells up to 17 cm thick are reported. The spectral flux above 3 MeV was measured using a liquid scintillation detector. At 17 cm thickness, also the total neutron multiplication was measured using a Bonner sphere system. The results agree well with calculations using beryllium nuclear data from the EFF-1 or the ENDF/B-Vi library. (author). 23 refs, 4 figs, 1 tab

  18. Benchmarking of the FENDL-3 Neutron Cross-Section Data Library for Fusion Applications

    International Nuclear Information System (INIS)

    Fischer, U.; Kondo, K.; Angelone, M.; Batistoni, P.; Villari, R.; Bohm, T.; Sawan, M.; Walker, B.; Konno, C.

    2014-03-01

    This report summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) with the objective to test and qualify the neutron induced general purpose FENDL-3.0 data library for fusion applications. The benchmark approach consisted of two major steps including the analysis of a simple ITER-like computational benchmark, and a series of analyses of benchmark experiments conducted previously at the 14 MeV neutron generator facilities at ENEA Frascati, Italy (FNG) and JAEA, Tokai-mura, Japan (FNS). The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses analysed. There is a slight trend, however, for an increase of the fast neutron flux in the shielding experiment and a decrease in the breeder mock-up experiments. The photon flux spectra measured in the bulk shield and the tungsten experiments are significantly better reproduced with FENDL-3.0 data. In general, FENDL-3, as compared to FENDL-2.1, shows an improved performance for fusion neutronics applications. It is thus recommended to ITER to replace FENDL-2.1 as reference data library for neutronics calculation by FENDL-3.0. (author)

  19. Role of inert gases in first wall phenomena in fusion devices

    International Nuclear Information System (INIS)

    Das, S.K.

    1979-01-01

    The first wall surfaces of fusion devices will be exposed to bombardment by inert gaseous projectiles such as helium. The flux, energy and angular distribution of the helium radiation will depend not only on the type of device but also on its design parameters. For near term tokamak devices, the first wall surface phenomena caused by helium bombardment that appear to be quite important are physical sputtering and radiation blistering. Examples of these processes for a number of first wall candidate materials are discussed. While the physical sputtering phenomen is well understood, the mechanism of blister formation is still not fully understood. The various models proposed for radiation blistering of metal during helium bombardment is critically reviewed in the light of most recent experimental results

  20. Protective interior wall and attaching means for a fusion reactor vacuum vessel

    International Nuclear Information System (INIS)

    Phelps, R.D.; Upham, G.A.; Anderson, P.M.

    1988-01-01

    A protective wall for the interior surface of a fusion reactor vessel wall is described comprising: an array of plates, each plate of the array including a main body section, a pair of edge sections bent at an angle with respect to the main body section, and a pair of flange-like end sections each having protruding sections with cut-aways therein, the protruding sections of the flange-like end sections extending in a direction substantially parallel to the main body section; and means operatively associated with the protruding sections of the flange-like end sections of the plates for mounting the array of plates to an associated vessel wall to be protected

  1. Simulations of fusion chamber dynamics and first wall response in a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR)

    Energy Technology Data Exchange (ETDEWEB)

    Qi, J.M., E-mail: qjm06@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Wang, Z., E-mail: wangz_es@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Chu, Y.Y., E-mail: chuyanyun@caep.cn [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China); Center for Fusion Energy Science and Technology (CFEST), China Academy of Engineering Physics, Mianyang 621999 (China); Li, Z.H., E-mail: lee_march@sina.com [Laboratory of Advanced Nuclear Energy (LANE), Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, Mianyang 621999 (China)

    2016-03-15

    Highlights: • Z-FFR utilizes DT neutrons to drive a sub-critical fission blanket to produce energy. • A metal shell and Ar gas are employed in the fusion chamber for shock mitigation. • Massive materials can effectively mitigate the thermal heats on the chamber wall. • The W-coated Zr-alloy first wall exhibits good viability as a long-lived component. - Abstract: In a Z-pinch driven fusion–fission hybrid power reactor (Z-FFR), the fusion target will produce enormous energy of ∼1.5 GJ per pulse at a frequency of 0.1 Hz. Almost 20% of the fusion energy yield, approximately 300 MJ, is released in forms of pulsed X-rays. To prevent the first wall from fatal damages by the intense X-rays, a thin spherical metal shell and rare Ar buffer gas are introduced to mitigate the transient X-ray bursts. Radiation hydrodynamics in the fusion chamber were investigated by MULTI-1D simulations, and the corresponding thermal and mechanical loads on the first wall were also obtained. The simulations indicated that by optimizing the design parameters of the metal shell and Ar buffer gas, peak power flux of the thermal heats on the first wall could be mitigated to less than 10{sup 4} W/cm{sup 2} within a time scale of several milliseconds, while peak overpressures of the mechanical loads varying from 0.6 to 0.7 MPa. In addition, the thermomechanical response in a W–coated Zr-alloy first wall was performed by FWDR1D calculations using the derived thermal and mechanical loads as inputs. The temperature and stress fields were analyzed, and the corresponding elastic strains were conducted for primary lifetime estimations by using the Coffin–Manson relationships of both W and Zr-alloy. It was shown that the maximum temperature rises and stresses in the first wall were less than 50 K and 130 MPa respectively, and lifetime of the first wall would be in excess of 10{sup 9} cycles. The chamber exhibits good viability as a long-lived component to sustain the Z-FFR conceptual

  2. A comparison of hydrogen vs. helium glow discharge effects on fusion device first-wall conditioning

    International Nuclear Information System (INIS)

    Dylla, H.F.

    1989-09-01

    Hydrogen- and deuterium-fueled glow discharges are used for the initial conditioning of magnetic fusion device vacuum vessels following evacuation from atmospheric pressure. Hydrogenic glow discharge conditioning (GDC) significantly reduces the near-surface concentration of simple adsorbates, such as H 2 O, CO, and CH 4 , and lowers ion-induced desorption coefficients by typically three orders of magnitude. The time evolution of the residual gas production observed during hydrogen-glow discharge conditioning of the carbon first-wall structure of the TFTR device is similar to the time evolution observed during hydrogen GDC of the initial first-wall configuration in TFTR, which was primarily stainless steel. Recently, helium GDC has been investigated for several wall-conditioning tasks on a number of tokamaks including TFTR. Helium GDC shows negligible impurity removal with stainless steel walls. For impurity conditioning with carbon walls, helium GDC shows significant desorption of H 2 O, CO, and CO 2 ; however, the total desorption yield is limited to the monolayer range. In addition, helium GDC can be used to displace hydrogen isotopes from the near-surface region of carbon first-walls in order to lower hydrogenic retention and recycling. 38 refs., 6 figs

  3. Thermonuclear fusion

    International Nuclear Information System (INIS)

    Weisse, J.

    2000-01-01

    This document takes stock of the two ways of thermonuclear fusion research explored today: magnetic confinement fusion and inertial confinement fusion. The basic physical principles are recalled first: fundamental nuclear reactions, high temperatures, elementary properties of plasmas, ignition criterion, magnetic confinement (charged particle in a uniform magnetic field, confinement and Tokamak principle, heating of magnetized plasmas (ohmic, neutral particles, high frequency waves, other heating means), results obtained so far (scale laws and extrapolation of performances, tritium experiments, ITER project), inertial fusion (hot spot ignition, instabilities, results (Centurion-Halite program, laser experiments). The second part presents the fusion reactor and its associated technologies: principle (tritium production, heat source, neutron protection, tritium generation, materials), magnetic fusion (superconducting magnets, divertor (role, principle, realization), inertial fusion (energy vector, laser adaptation, particle beams, reaction chamber, stresses, chamber concepts (dry and wet walls, liquid walls), targets (fabrication, injection and pointing)). The third chapter concerns the socio-economic aspects of thermonuclear fusion: safety (normal operation and accidents, wastes), costs (costs structure and elementary comparison, ecological impact and external costs). (J.S.)

  4. Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

    1976-05-01

    The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

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

    International Nuclear Information System (INIS)

    Ma Xubo; Chen Yixue; Gao Bin

    2013-01-01

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

  6. DT fusion neutron irradiation of BNL--LASL superconductor wires

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1976-01-01

    The following samples were irradiated with the LLL rotating target neutron source: 19-core Nb 3 Sn multifilament wires, Nb 3 Sn single core, V 3 Ga single core, NbTi Supercon 402, and NbTi cupronickel jacketed. No test results are given

  7. Evaluation of light nuclides for fusion neutronics application

    International Nuclear Information System (INIS)

    Shibata, Keiichi

    1984-01-01

    Neutron nuclear data of 6 Li, 7 Li, 12 C and 16 O have been evaluated for JENDL-3. The methods of evaluation and the results are described. The present evaluated data are employed for the analysis of the Japan-US co-operative experiment using FNS. (author)

  8. Results of strategic calculations for optimizing the first wall life in a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Daenner, W.

    1981-01-01

    The development of the FWLTB computer program has reached a stage where prediction of the first wall lifetime is possible. Because of the large number of free parameters strategic calculations were found to be the most appropriate way to arrive at load design conditions which allow optimum life expectancy. In this paper a revised set of life criteria is presented this being followed by the results of parameter studies in which single parameters were varied while the remaining ones were kept fixed at a reference value. These results are used as a guide during the subsequent strategic calculations. In a first strategy we aimed at finding the maximum lifetime for the case that the reactor is operated at a neutron wall loading of 10 MW/m 2 . We found that operation over a period of more than one year is possible if the first wall is designed in a very tiny geometry and cooled by a low-pressure coolant. In a second strategy the aim was to find the design conditions for the case that the first wall is cooled by a high-pressure coolant. It is shown that liquid-lithium cooling is manageable up to high wall loadings, but the lifetime is restricted to about 6 MWa/m 2 . Helium cooling allows a higher lifetime, but the design conditions are such that only modest wall loadings can be permitted. (orig.)

  9. An electrically conducting first wall for the fusion engineering device-A (FED-A) tokamak

    International Nuclear Information System (INIS)

    Cramer, B.A.; Fuller, G.M.

    1983-01-01

    The first wall of the tokamak FED-A device was designed to satisfy two conflicting requirements. They are a low electrical resistance to give a long eddy-current decay time and a high neutron transparency to give a favorable tritium breeding ratio. The tradeoff between these conflicting requirements resulted in a copper alloy first wall that satisfied the specific goals for FED-A, i.e., a minimum eddy-current decay time of 0.5 sec and a tritium breeding ratio of at least 1.2. Aluminum alloys come close to meeting the requirements and would also probably work. Stainless steel will not work in this application because shells thin enough to satisfy temperature and stress limits are not thick enough to give a long eddy-current decay time and to avoid disruption induced melting. The baseline first wall design is a rib-stiffened, double-wall construction. The total wall thickness is 1.5 cm, including a water coolant thickness of 0.5 cm. The first wall is divided into twelve 30-degree sectors. Flange rings at the ends of each sector are bolted together to form the torus. Structural support is provided at the top center of each sector

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

    Science.gov (United States)

    Shi, Xue-Ming; Peng, Xian-Jue

    2016-09-01

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

  11. A GDT-based fusion neutron source for academic and industrial applications

    Science.gov (United States)

    Anderson, J. K.; Forest, C. B.; Mirnov, V. V.; Peterson, E. E.; Waleffe, R.; Wallace, J.; Harvey, R. W.

    2017-10-01

    The design of a fusion neutron source based on the gas dynamic trap (GDT) configuration is underway. The motivation is both the ends and the means. There are immediate applications for neutrons including medical isotope production and actinide burners. Taking the next step in the magnetic mirror path will leverage advances in high-temperature superconducting magnets and additive manufacturing in confining a fusion plasma, and both the technological and physics bases exist. Recent breakthrough results at the GDT facility in Russia demonstrate stable confinement of a beta 60% mirror plasma at high Te ( 1 keV). These scale readily to a fusion neutron source with an increase in magnetic field, mirror ratio, and ion energy. Studies of a next-step compact device focus on calculations of MHD equilibrium and stability, and Fokker-Planck modeling to optimize the heating scenario. The conceptualized device uses off-the-shelf MRI magnets for a 1 T central field, REBCO superconducting mirror coils (which can currently produce fields in excess of 30T), and existing 75 keV NBI and 140 GHz ECRH. High harmonic fast wave injection is damped on beam ions, dramatically increasing the fusion reactivity for an incremental bump in input power. MHD stability is achieved with the vortex confinement scheme, where a biasing profile imposes optimal ExB rotation of the plasma. Liquid metal divertors are being considered in the end cells. Work supported by the Wisconsin Alumni Research Foundation.

  12. Reducing risk and accelerating delivery of a neutron source for fusion materials research

    Energy Technology Data Exchange (ETDEWEB)

    Surrey, E., E-mail: Elizabeth.Surrey@ccfe.ac.uk [EURATOM/CCFE, Abingdon OX14 3DB (United Kingdom); Porton, M. [EURATOM/CCFE, Abingdon OX14 3DB (United Kingdom); Davenne, T.; Findlay, D.; Letchford, A.; Thomason, J. [STFC Rutherford Appleton Laboratory, Harwell OX11 0QX (United Kingdom); Roberts, S.G.; Marrow, J.; Seryi, A. [University of Oxford, Oxford OX1 3DP (United Kingdom); Connolly, B. [University of Birmingham, Edgbaston B15 2TT (United Kingdom); Owen, H. [University of Manchester, Manchester M13 9PL (United Kingdom)

    2014-04-15

    Highlights: • Proposed neutron source for fusion materials – FAFNIR – n(d,C) stripping source. • Near term technology, reduces risk compared with IFMIF, timely data production. • Technical, economic and programme needs assessed, compatible with EU Roadmap proposals. • Safety case impacts regulatory role for source, now mainly stakeholder insurance. - Abstract: The materials engineering database relevant to fusion irradiation is poorly populated and it has long been recognized that a fusion spectrum neutron source will be required, the facility IFMIF being the present proposal. Re-evaluation of the regulatory approach for the EU proposed DEMO device shows that the purpose of the source can be changed from lifetime equivalent irradiation exposure to data generation at lower levels of exposure by adopting a defence in depth strategy and regular component surveillance. This reduces the specification of the source with respect to IFMIF allowing lower risk technology solutions to be considered. A description of such a source, the Facility for Fusion Neutron Irradiation Research, FAFNIR, is presented here along with project timescales and costs.

  13. Falling liquid film flow along cascade-typed first wall of laser-fusion reactor

    International Nuclear Information System (INIS)

    Kunugi, T.; Nakai, T.; Kawara, Z.

    2007-01-01

    To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a 'cascade-typed' first wall with a falling liquid film flow is proposed as the 'liquid wall' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the STREAM code and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ρu 2 δ/σ: ρ is density, u is velocity, δ is film thickness, σ is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant water-head located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same structure and the same height as the reactor design

  14. DD fusion neutron production at UW-Madison using IEC devices

    Science.gov (United States)

    Fancher, Aaron; Michalak, Matt; Kulcinski, Gerald; Santarius, John; Bonomo, Richard

    2017-10-01

    An inertial electrostatic confinement (IEC) device using spherical, gridded electrodes at high voltage accelerates deuterium ions, allowing for neutrons to be produced within the device from DD fusion reactions. The effects of the device cathode voltage (30-170 kV), current (30-100 mA), and pressure (0.15-1.25 mTorr) on the neutron production rate have been measured. New high voltage capabilities have resulted in the achievement of a steady state neutron production rate of 3.3x108 n/s at 175 kV, 100 mA, and 1.0 mTorr of deuterium. Applications of IEC devices include the production of DD neutrons to detect chemical explosives and special nuclear materials using active interrogation methods. Research supported by US Dept. of Homeland Security Grant 2015-DN-077-AR1095 and the Grainger Foundation.

  15. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  16. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2001-01-01

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

  17. Measurement and Analysis of Activation Induced in Lanthanum, Erbium and Tantalum by Fusion Peak Neutrons

    International Nuclear Information System (INIS)

    Klix, A.; Eichin, R.; Freiesleben, H.; Schomburg, K.; Seidel, K.; Unholzer, S.; Forrest, R.A.

    2006-01-01

    The large fluxes of neutrons in the materials of a fusion device during operation produce activation that is relevant to operational safety and decommissioning. Nuclides with a broad range of half-lives have to be included in the corresponding analyses. The activity with decay times ranging from the order of magnitude of minutes to weeks is of interest with respect to heat production and shut-down dose rates, whereas the long-term activity determines the waste management. The activity is mainly produced by two components of the neutron flux spectrum, by thermal neutrons and by the 14-MeV D-T fusion neutrons. Analyses of the material activation rely on calculations with inventory codes and libraries containing activation and decay data. To gain trust in the results of such calculations data and codes have to be validated experimentally. In the present work, the European Activation System (EASY, inventory code FISPACT and data library EAF) was tested in benchmark experiments on Lanthanum, Erbium and Tantalum. They are constituents of fusion reactor structural materials such as EUROFER and insulating coatings for liquid breeder systems. Small samples of the materials were irradiated in a D-T neutron field. The gamma-radioactivity following irradiation was measured several times during decay and nuclide activities were derived. For each of the measured activities the corresponding value was calculated with EASY, and the calculated-to-experimental ratios (C/E) were determined. The nuclear reactions producing the activities were also analysed. The C/E ratios obtained for the individual activities will be used for discussing the activation performance and the contact dose rate of the materials at fusion power plant conditions. (author)

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

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

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

  19. Estimation of the {alpha} particles and neutron distribution generated during a fusion reaction; Evaluation de la distribution des particules {alpha} et des neutrons issus de la reaction de fusion

    Energy Technology Data Exchange (ETDEWEB)

    Dellacherie, S.

    1997-12-01

    The respective distributions (or density probabilities) of {alpha} particles and neutrons have been modeled using a Monte-Carlo method for the thermonuclear fusion reaction D + T {yields} {alpha} + n + 17.6 MeV. (N.T.).

  20. The plasma-wall interaction region: a key low temperature plasma for controlled fusion

    International Nuclear Information System (INIS)

    Counsell, G F

    2002-01-01

    The plasma-wall interaction region of a fusion device provides the interface between the hot core plasma and the material surfaces. To obtain acceptably low levels of erosion from these surfaces requires most of the power leaving the core to be radiated. This is accomplished in existing devices by encouraging plasma detachment, in which the hot plasma arriving in the region is cooled by volume recombination and ion-neutral momentum transfer with a dense population of neutrals recycled from the surface. The result is a low temperature (1 eV e e >10 19 m -3 ) but weakly ionized (n 0 >10 20 m -3 , n e /n 0 <0.1) plasma found nowhere else in the fusion environment. This plasma provides many of the conditions found in industrial plasmas exploiting plasma chemistry and the presence of carbon in the region (in the form of carbon-fibre composite used in the plasma facing materials) can result in the formation of deposited hydrocarbon films. The plasma-wall interaction region is therefore among the most difficult in fusion to model, requiring an understanding of atomic, molecular and surface physics issues

  1. Neutron induced displacement damage in beryllium in the blanket of a (d,t)-fusion reactor

    International Nuclear Information System (INIS)

    Hermanutz, D.

    1995-09-01

    Beryllium is a favoured candidate for a neutron multiplier in solid breeder blankets of fusion reactors. This is mainly due to its low (n, 2n)-reaction threshold and because of its good thermal and mechanical properties. Its behaviour under intense neutron irradiation, however, is a crucial issue for its use in future fusion reactors. Displacement damage in beryllium so far has been calculated both with data related and methodological deficiencies. First of all, there is a need to have accurate cross-section data in order to obtain reliable spectra of primary knock-on atoms (PKA's). Furthermore, there are principal restrictions of the NRT-model in general used to calculate secondary displacements initiated by PKA's. The underlying theory of damage-energy (part of kinetic energy of PKA transferred elastically to matrix atoms) according to Lindhard is strictly valid only for medium and heavy mass ions with moderate energies in targets of the same element. In this work improved damage cross-sections and displacement rates (dpa/s) in beryllium have been calculated based on cross-section data from ENDF/B-VI (with a significantly improved (n, 2n)-evaluation) and on an appropriate treatment of damage-energy that is suitable for fusion relevant damage of light mass materials. ''This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Communities within the European Fusion Technology Program''. (orig.)

  2. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2000-01-01

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

  3. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2000-07-01

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

  4. Vessel wall damage by X-rays and 15 MeV neutrons

    International Nuclear Information System (INIS)

    Aarnoudse, M.W.

    1979-01-01

    In two simple mucopolysaccharide systems, synovial fluid and subcutaneous connective tissue membranes, the degrading effects of 200 kVp X-rays and 15 MeV neutrons is compared. Due to the depolymerization of the mucopolysaccharides the viscosity of synovial fluid decreases and the permeability of the connective tissue membranes for saline increases after irradiation. In both systems a RBE of 0.6 has been found for fast neutrons. The atheromatous changes in the wall of elastic arteries (lipid penetration into the vessel wall and the formation of plaques consisting of large, lipid-filled foam cells) are studied in the carotid arteries of hypercholesterolemic rabbits, two months after irradiating the arteries with different doses of X-rays or neutrons. (Auth.)

  5. Pre-compound neutron evaporation in heavy ion fusion reactions

    International Nuclear Information System (INIS)

    Ajay Kumar; Singh, Hardev; Rajesh Kumar; Govil, I.M.; Golda, K.S.; Rakesh Kumar; Datta, S.K.; Yogi, B.K.; Viesti, G.

    2007-01-01

    In the present investigation, exclusive neutron spectra have been measured in coincidence with the observed γ-rays characteristic to the residual nuclei 53 Fe, 55 Fe and 56 Co. In this new experimental study, we have followed the strategy of forming the compound nucleus 58 Ni at the same excitation energy through two different entrance channels viz the mass symmetric reaction 31 P + 27 Al and the mass asymmetric reaction 12 C + 46 Ti

  6. Measurement and analysis of 14 MeV neutron-induced double-differential neutron emission cross sections needed for fission and fusion reactor technology

    International Nuclear Information System (INIS)

    Wang Dahai.

    1990-10-01

    The main objectives of this IAEA Co-ordinated Research Programme are to improve the data on 14 MeV neutron-induced double-differential neutron emission cross sections for materials needed for fission and fusion reactor technology. This report summarizes the conclusions and recommendations which were agreed by all participants during the Second Research Co-ordination Meeting

  7. Neutron degradation of UV enhanced optical fibers for fusion installation plasma diagnostics

    International Nuclear Information System (INIS)

    Sporea, D.; Vata, I.; Dudu, D.; Danis, Ana

    2004-01-01

    The remote diagnostics of plasmas in fusion installations requires adequate connection links to transfer the measured signals in media subjected to high electromagnetic disturbances. We evaluated the neutron irradiation induced optical absorption in several commercially available optical fibers, as they were assessed for their possible use in fusion installations. Optical fiber samples were subjected to subsequent irradiation with fluences from 6 x 10 11 to 6 x 10 13 n/cm 2 . Significant radiation induced absorption was observed in the 220-260 nm spectral band, mainly for small core diameter optical fibers (200/220 μm), independent of the cladding material used

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  9. Recent developments in neutron dosimetry and radiation damage calculations for fusion-materials studies

    International Nuclear Information System (INIS)

    Greenwood, L.R.

    1983-01-01

    This paper is intended as an overview of activities designed to characterize neutron irradiation facilities in terms of neutron flux and energy spectrum and to use these data to calculate atomic displacements, gas production, and transmutation during fusion materials irradiations. A new computerized data file, called DOSFILE, has recently been developed to record dosimetry and damage data from a wide variety of materials test facilities. At present data are included from 20 different irradiations at fast and mixed-spectrum reactors, T(d,n) 14 MeV neutron sources, Be(d,n) broad-spectrum sources, and spallation neutron sources. Each file entry includes activation data, adjusted neutron flux and spectral data, and calculated atomic displacements and gas production. Such data will be used by materials experimenters to determine the exposure of their samples during specific irradiations. This data base will play an important role in correlating property changes between different facilities and, eventually, in predicting materials performance in fusion reactors. All known uncertainties and covariances are listed for each data record and explicit references are given to nuclear decay data and cross sections

  10. Development of Portable Pulsed Neutron Generators Utilizing a D-T or D-D Fusion Reaction

    International Nuclear Information System (INIS)

    Nishimura, Kazuya; Miake, Yoshinobu; Kato, Michio; Rintsu, Yukou

    2001-01-01

    Prototypes of sealed neutron tubes in a D-T or D-D fusion reaction for logging while drilling (LWD) were developed; then operational tests were performed to check their functional properties. One of the prototypes passed most of the specified conditions for using LWD. Further studies were needed to put a sealed neutron tube into practical use. For applications to other fields, such as an in situ calibration source for neutron detector efficiencies and an in situ calibration source for fusion systems, a sealed neutron tube is needed to have higher-intensity neutron output and a long life. Thus, the performance of the ion source used in the neutron tube is improved to obtain high gas utilization efficiencies or low-pressure operation with high ionization efficiencies. The characteristics of the new ion sources used in the foregoing sealed neutron tube are discussed in terms of preliminary tests. The aforementioned performances are obtained

  11. Evaluation of the fusion-related neutron nuclear data for JENDL-3

    International Nuclear Information System (INIS)

    Chiba, Satoshi

    1988-01-01

    Status of the neutron nuclear data evaluations for JENDL-3 will be described for nuclides important in the development of D-T fusion reactors. In this article, however, only explanation of the evaluations for the very light mass region will be presented to avoid overlapping with what are given in another papers submitted to this seminar. Emphases are placed on the tritium production cross sections, inelastic scattering cross sections including the double-differential neutron emission spectrum (DDX), threshold reaction cross sections and photon production cross sections. The methods employed to prepare JENDL-3T library and their results will be summarized. (author)

  12. FENDL/A-2.0. Neutron activation cross section data library for fusion applications

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.; Kopecky, J.; Sublet, J.C. Sublet; Forrest, R.A.

    1997-01-01

    This document describes the contents of a comprehensive neutron cross section data library for 13,006 neutron activation reactions with 739 target nuclides from H (A=1,Z=1) to Cm (A=248,Z=96), in the incident energy range up to 20 MeV. FENDL/A-2 is a sublibrary of FENDL-2, the second revision of the evaluated nuclear data library for fusion applications. It is supplemented by a decay data library FENDL/D-2 in ENDF-6 format for 1867 nuclides. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape upon request. (author)

  13. Analysis of the neutron time-of-flight spectra from inertial confinement fusion experiments

    Energy Technology Data Exchange (ETDEWEB)

    Hatarik, R., E-mail: hatarik1@llnl.gov; Sayre, D. B.; Caggiano, J. A.; Phillips, T.; Eckart, M. J.; Bond, E. J.; Cerjan, C.; Grim, G. P.; Hartouni, E. P.; Mcnaney, J. M.; Munro, D. H. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Knauer, J. P. [Laboratory for Laser Energetics, University of Rochester, Rochester, New York 14623 (United States)

    2015-11-14

    Neutron time-of-flight diagnostics have long been used to characterize the neutron spectrum produced by inertial confinement fusion experiments. The primary diagnostic goals are to extract the d + t → n + α (DT) and d + d → n + {sup 3}He (DD) neutron yields and peak widths, and the amount DT scattering relative to its unscattered yield, also known as the down-scatter ratio (DSR). These quantities are used to infer yield weighted plasma conditions, such as ion temperature (T{sub ion}) and cold fuel areal density. We report on novel methodologies used to determine neutron yield, apparent T{sub ion}, and DSR. These methods invoke a single temperature, static fluid model to describe the neutron peaks from DD and DT reactions and a spline description of the DT spectrum to determine the DSR. Both measurements are performed using a forward modeling technique that includes corrections for line-of-sight attenuation and impulse response of the detection system. These methods produce typical uncertainties for DT T{sub ion} of 250 eV, 7% for DSR, and 9% for the DT neutron yield. For the DD values, the uncertainties are 290 eV for T{sub ion} and 10% for the neutron yield.

  14. Neutron time-of-flight counters and spectrometers for diagnostics of burning fusion plasmas

    International Nuclear Information System (INIS)

    Elevant, T.; Olsson, M.

    1991-02-01

    Experiment with burning fusion plasmas in tokamaks will place particular requirements on neutron measurements from radiation resistance-, physics-, burn control- and reliability considerations. The possibility to meet these needs by measurements of neutron fluxes and energy spectra by means of time-of-flight techniques are described. Reference counters and spectrometers are proposed and characterized with respect to efficiency, count-rate capabilities, energy resolution and tolerable neutron and γ-radiation background levels. The instruments can be used in a neutron camera and are capable to operate in collimated neutron fluxes up to levels corresponding to full nuclear output power in the next generation of experiments. Energy resolutions of the spectrometers enables determination of ion temperatures from 3 (keV) through analysis of the Doppler broadening. Primarily, the instruments are aimed for studies of 14 (MeV) neutrons produced in (d,t)-plasmas but can, after minor modifications, be used for analysis of 2.45 (MeV) neutrons produced in (d,d)-plasma. (au) (33 refs.)

  15. Effect of laser spot size on fusion neutron yield in laser–deuterium cluster interactions

    International Nuclear Information System (INIS)

    Chen Guanglong; Lu Haiyang; Wang Cheng; Liu Jiansheng; Li Ruxin; Ni Guoquan; Xu Zhizhan

    2008-01-01

    The effect of the laser spot size on the neutron yield of table-top nuclear fusion from explosions of a femtosecond intense laser pulse heated deuterium clusters is investigated by using a simplified model, in which the cluster size distribution and the energy attenuation of the laser as it propagates through the cluster jet are taken into account. It has been found that there exists a proper laser spot size for the maximum fusion neutron yield for a given laser pulse and a specific deuterium gas cluster jet. The proper spot size, which is dependent on the laser parameters and the cluster jet parameters, has been calculated and compared with the available experimental data. A reasonable agreement between the calculated results and the published experimental results is found

  16. Large area imaging of hydrogenous materials using fast neutrons from a DD fusion generator

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, J.T., E-mail: ted@adelphitech.com [Adelphi Technology Inc., 2003 East Bayshore Road, Redwood City, California 94063 (United States); Williams, D.L.; Gary, C.K.; Piestrup, M.A.; Faber, D.R.; Fuller, M.J.; Vainionpaa, J.H.; Apodaca, M. [Adelphi Technology Inc., 2003 East Bayshore Road, Redwood City, California 94063 (United States); Pantell, R.H.; Feinstein, J. [Department of Electrical Engineering, Stanford University, Stanford, California 94305 (United States)

    2012-05-21

    A small-laboratory fast-neutron generator and a large area detector were used to image hydrogen-bearing materials. The overall image resolution of 2.5 mm was determined by a knife-edge measurement. Contact images of objects were obtained in 5-50 min exposures by placing them close to a plastic scintillator at distances of 1.5 to 3.2 m from the neutron source. The generator produces 10{sup 9} n/s from the DD fusion reaction at a small target. The combination of the DD-fusion generator and electronic camera permits both small laboratory and field-portable imaging of hydrogen-rich materials embedded in high density materials.

  17. Neutron irradiation effects on superconducting and stabilizing materials for fusion magnets

    International Nuclear Information System (INIS)

    Maurer, W.

    1984-05-01

    Available low-temperature neutron irradiation data for the superconductors NbTi and Nb 3 Sn and the stabilization materials Cu and Al are collected and maximum tolerable doses for these materials are defined. A neutron flux in a reactor of about 10 9 n/cm 2 s at the magnet position is expected. However, in fusion experiments the flux can be higher by an order of magnitude or more. The energy spectrum is similar to a fission reactor. A fluence of about 10 18 n/cm 2 results during the lifetime of a fusion magnet (about 20 full power years). At this fluence and energy spectrum no severe degradation of the superconducting properties of NbTi and Nb 3 Sn will occur. But the radiation-induced resistivity is for Cu about a twentieth of the room temperature resistivity and a tenth for Al. (orig.) [de

  18. Design and fabrication of foam-insulated cryogenic target for wet-wall laser fusion reactor

    International Nuclear Information System (INIS)

    Norimatsu, T.; Takeda, T.; Nagai, K.; Mima, K.; Yamanaka, T.

    2003-01-01

    A foam insulated cryogenic target was proposed for use in a future laser fusion reactor with a wet wall. This scheme can protect the solid DT layer from melting due to surface heating by adsorption of metal vapor without significant reduction in the target gain. Design spaces for the injection velocity and the acceptable vapor pressure in the reactor are discussed. Basic technology to fabricate such structure was demonstrated by emulsion process. Concept of a cryogenic fast-ignition target with a gold guiding cone was proposed together with direct injection filling of liquid DT. (author)

  19. The effect of optimal wall loads and blanket technologies on the cost of fusion electricity

    International Nuclear Information System (INIS)

    Knight, P.J.; Ward, D.J.

    2000-01-01

    This paper presents a discussion of trends in fusion economics based on technology, as well as, physics arguments. Based on relatively simple physics considerations, supported by detailed systems code calculations, it is shown that optimal wall loads are not high. The results of systems code calculations, focussing on the economic impact of different blanket technologies, are described. These suggest that the economically favourable thermodynamic efficiencies of some blankets capable of operating at higher temperatures may be counterbalanced by the economic penalties of shorter lifetimes

  20. Neutron emission as a probe of fusion-fission and quasi-fission dynamics

    International Nuclear Information System (INIS)

    Hinde, D.J.

    1991-01-01

    Pre- and post scission neutron yeilds have been measured as a function of projectile mass, compound nucleus fissility, and fission mass-split and total kinetic energy (TKE) for 27 fusion-fission and quasi-fission reactions induced by beams of 16,18 O, 40 Ar and 64 Ni. A new method of interpretation of experimental pre-scission neutron multiplicities ν-pre and mean kinetic energies ε ν allows the extraction of fission time scales with much less uncertainty than previously, all fusion-fission results being consistent with a dynamical time scale of (35±15) x 10 -21 s for symmetric fission. All reactions show that ν-pre falls quite rapidly with increasing mass-asymmetry; evidence is presented that for fusion-fission reactions this is partly due to a reduction of the dynamical fission time scale with mass-asymmetry. For quasi-fission, the data indicate that the pre-scission multiplicity and mean neutron kinetic energy are very sensitive to the final mass-asymmetry, but that the time scale is virtually independent of mass-asymmetry. It is concluded that for fusion-fission there is no dependence of ν-pre on TKE, whilst for 64 Ni-induced quasi-fission reactions, a strong increase of ν-pre with decreasing TKE is observed, probably largely caused by neutron emission during the acceleration time of the fission fragments in these fast reactions. Interpretation of post-scission multiplicities in terms of fragment excitation energies leads to deduced time scales consistent with those determined from the pre-scission data. 54 refs., 17 tabs., 25 figs

  1. Nuclear fusion and carbon flashes on neutron stars

    International Nuclear Information System (INIS)

    Taam, R.E.; Picklum, R.E.

    1978-01-01

    The properties of nuclear burning shells in the envelopes of accreting neutron stars are investigated for neutron star masses of 0.56M/sub sun/ and 1.41M/sub sun/ and mass accretion rates M ranging from 10 -11 M/sub sun/ yr -1 to 2 x 10 -9 M/sub sun/ yr -1 . It is found that (1) the hydrogen-burning shells lie at high density, log rhoapprox.6, (2) the hydrogen and helium shells overlap for M> or approx. =3 x 10 -10 M/sub sun/ yr -1 , and (3) the carbon abundance at the base of the helium shell is a strong function of M, being greater than 0.95 (less than 0.3) for less than 10 -10 M/sub sun/ yr -1 (greater than 10 -9 M/sub sun/ yr -1 ). A stability analysis of the hydrogen and helium burning shells reveals them to be unstable whenever they overlap. Detailed calculations of the thermal evolution of the carbon shells show that carbon flashes occur for 10 -10 -1 ) -9 . Results for lower rates are inconclusive

  2. Nuclear fusion and carbon flashes on neutron stars

    Science.gov (United States)

    Taam, R. E.; Picklum, R. E.

    1978-01-01

    This paper reports on detailed calculations of the thermal evolution of the carbon-burning shells in the envelopes of accreting neutron stars for mass-accretion rates of 1 hundred-billionth to 2 billionths of a solar mass per yr and neutron-star masses of 0.56 and 1.41 solar masses. The work of Hansen and Van Horn (1975) is extended to higher densities, and a more detailed treatment of nuclear processing in the hydrogen- and helium-burning regions is included. Results of steady-state calculations are presented, and results of time-dependent computations are examined for accretion rates of 3 ten-billionths and 1 billionth of solar mass per yr. It is found that two evolutionary sequences lead to carbon flashes and that the carbon abundance at the base of the helium shell is a strong function of accretion rate. Upper limits are placed on the accretion rates at which carbon flashes will be important.

  3. A preliminary model for estimating the first wall lifetime of a fusion reactor

    International Nuclear Information System (INIS)

    Daenner, W.

    1975-02-01

    The estimation of the first wall lifetime is a necessary basis for predicting the availability of a fusion power plant. In order to do this, an analytical model was prepared and programmed for the computer which calculates the temperature and stress load of the first wall from the principal design parameters and quotes them against the relevant material properties. Neither the analytical model nor the information about the material performance is yet complete so that the answers obtained from the program are very preliminary. This situation is underlined by the results of sample calculations performed for the CTRD blanket module cell. The results obtained for vanadium and vanadium alloys show a strong dependence of the lifetime on the irradiation creep and the ductility of these materials. Completion of this model is envisaged as soon as the missing information becomes available. (orig.) [de

  4. Effects of breakup of weakly bound projectile and neutron transfer on fusion reactions around Coulomb barrier

    International Nuclear Information System (INIS)

    Lin, C.J.; Zhang, H.Q.; Yang, F.; Ruan, M.; Liu, Z.H.; Wu, Y.W.; Wu, X.K.; Zhou, P.; Zhang, C.L.; Zhang, G.L.; An, G.P.; Jia, H.M.; Xu, X.X.

    2007-01-01

    The excitation functions of quasielastic and elastic scattering at backward angles have been measured for the systems of 16 O+ 152 Sm, 6,7 Li+ 208 Pb and 32 S+ 90,96 Zr. The barrier distributions are extracted from these measured excitation functions and compared with the corresponding fusion barrier distributions. Except some details, the barrier distributions derived from the data of fusion and quasielastic/elastic scattering are almost the same for the tightly bound reaction systems. For the reaction systems with weakly bound projectile, the barrier distributions extracted from quasielastic scattering are obviously different from the fusion barrier distributions. However, the barrier distributions extracted from the excitation functions of the quasielastic scattering plus breakup are almost the same as the one extracted from the complete fusion data. This result means that barrier distribution not only bears the information of nuclear structures but also contains the knowledge of reaction mechanisms. Our results show that the complete fusion of the weakly bound projectile with heavy target is suppressed at the above barrier energies as compared with the model predictions. In addition, the measured barrier distribution of 32 S+ 96 Zr is broaden and extends to lower energy than in the case of 32 S+ 90 Zr due to the coupling of neutron transfer with positive Q-values, which result in a significant enhancement of fusion cross sections at the subbarrier energies

  5. Laboratory feasibility study of fusion vessel inner wall chemical analysis by Laser Induced Breakdown Spectroscopy

    International Nuclear Information System (INIS)

    Almaviva, Salvatore; Caneve, Luisa; Colao, Francesco; Fantoni, Roberta; Maddaluno, Giorgio

    2012-01-01

    Graphical abstract: Laser-Induced-Breakdown-Spectroscopy was used for the determination of the atomic composition of multilayered samples simulating the tiles of plasma facing components in the next generation fusion machines. Highlights: ► Description and characterization of an LIBS set-up for diagnostics in fusion machines. ► Identification of atomic composition of multilayered tiles simulating plasma facing components. ► Qualitative applicability of the Calibration Free method for quantitative analysis. ► Feasibility of large scale application in the processes of control during the tiles fabrication. ► Feasibility of erosion monitoring during operation of fusion machines. - Abstract: Laser Induced Breakdown Spectroscopy (LIBS) is nowadays a well established tool for qualitative, semi-quantitative and quantitative analyses of surfaces, with micro-destructive characteristics and capabilities for stratigraphy. LIBS is an appealing technique compared with many other types of elemental analysis thanks to the set up versatility facilitating non-invasive and remote analyses, as well as suitability to diagnostics in harsh environments. In this work, LIBS capabilities were used for the determination of the atomic composition of multilayered samples simulating the tiles of plasma facing components in the next generation fusion machines such as ITER. A new experimental setup was designed and realized in order to optimize the characteristics of an LIBS system working at low pressure and remotely, as it should be for an in situ system to be applied in monitoring the erosion and redeposition phenomena occurring on the inner walls of a fusion device. The effects of time delay and laser fluence on LIBS sensitivity at reduced pressure were examined, looking for operational conditions suitable to analytical applications. The quantitative analysis of some atomic species in the superficial layer has been carried out using a Calibration Free (CF) approach in the time

  6. Measurements of strongly localized potential well profiles in an inertial electrostatic fusion neutron source

    International Nuclear Information System (INIS)

    Yoshikawa, K.; Takiyama, K.; Koyama, T.

    2001-01-01

    Direct measurements of localized electric fields are made by the laser-induced fluorescence (LIF) method by use of the Stark effects in the central cathode core region of an Inertial-Electrostatic Confinement Fusion (IECF) neutron (proton) source, which is expected for various applications, such as luggage security inspection, non-destructive testing, land mine detector, or positron emitter production for cancer detection, currently producing continuously about 10 7 n/sec D-D neutrons. Since 1967 when the first fusion reaction was successfully proved experimentally in a very compact IECF device, potential well formation due to space charge associated with spherically converging ion beams has been a central key issue to be clarified in the beam-beam colliding fusion, which is the major mechanism of the IECF neutron source. Many experiments, but indirect, were made so far to clarify the potential well, but none of them produced definitive evidence, however. Results by the present LIF method show a double well potential profile with a slight concave for ion beams with relatively larger angular momenta, whereas for ions with smaller angular momenta, potential but much steeper peak to develop. (author)

  7. Activation analysis of tritium breeder lithium lead irradiated by fusion neutrons in FDS-II

    International Nuclear Information System (INIS)

    Mingliang Chen

    2006-01-01

    R-and-D of fusion materials, especially their activation characteristics, is one of the key issues for fusion research in the world. Research on activation characteristics for low activation materials, such as reduced activation ferritic/martensitic steels, vanadium alloys and SiCf/SiC composites, is being done throughout the world to ensure the attractiveness of fusion power regarding safety and environmental aspects. However, there is less research on the activation characteristics of the other important fusion materials, such as tritium breeder etc.. Lithium lead (Li 17 Pb 83 ) is presently considered as a primary candidate tritium breeder for fusion power reactors because of its attractive characteristics. It can serve as a tritium breeder, neutron multiplier and coolant in the blanket at the same time. The radioactivity of Li 17 Pb 83 by D-T fusion neutrons in FDS-II has been calculated and analyzed. FDS-II is a concept design of fusion power reactor, which consists of fusion core with advanced plasma parameters extrapolated from the ITER (International Thermonuclear Experimental Reactor) and two candidates of liquid lithium breeder blankets (named SLL and DLL blankets). The neutron transport and activation calculation are carried out based on the one-dimensional model for FDS-II with the home-developed multi-functional code system VisualBUS and the multi-group data library HENDL1.0/MG and European Activation File EAF-99. The effects of irradiation time on the activation characteristics of Li 17 Pb 83 were analyzed and it concludes that the irradiation time has an important effect on the activation level of Li 17 Pb 83 . Furthermore, the results were compared with the activation levels of other tritium breeders, such as Li 4 SiO 4 , Li 2 TiO 3 , Li 2 O and Li etc., under the same irradiation conditions. The dominant nuclides to dose rate and activity of Li 17 Pb 83 were analyzed as well. Tritium generated by Li has a great contribution to the afterheat and

  8. Neutron shield analysis and design for the PDX fusion facility

    International Nuclear Information System (INIS)

    Grimesey, R.A.; Nigg, D.W.; Scott, A.J.; Wheeler, F.J.; Jassby, D.L.; Perry, E.D.

    1979-01-01

    The basic component of the biological shield for PDX is an existing 81 cm thick high-density concrete shielding wall surrounding the machine. The principal additional shielding requirement is a roof shield over the machine to reduce air-scattered skyshine dose into the PDX control room and to the site boundary. The roof shield is designed in removable sections on a steel support structure permitting overhead crane access to major PDX components. After analysis of a number of alternate concepts, a roof shield consisting of 50 cm of water in polyethylene tanks was selected to meet design objectives of effectiveness, weight, removability, and cost

  9. Determination of procedures for transmutation of fission product wastes by fusion neutrons. Volume 2. Final report

    International Nuclear Information System (INIS)

    Lang, G.P.

    1980-12-01

    This study is concerned with the engineering aspects of the transmutation of fission products utilizing neutrons generated in fusion reactors. It is assumed that fusion reactors, although not yet developed, will be available around the turn of the century. Therefore, early studies of this type are appropriate as a guide to the large amount of further investigations that will be needed to fully evaluate this concept. Not all of the radioactive products from light water reactors can be economically transmuted, but it appears that the most hazardous can. This requires that fission-product wastes must first be separated into a number of fractions, and in some instances this must be accomplished with extremely high separation factors. A review of current commercial separation processes and of promising methods that are now in the laboratory stage indicate that the necessary processes can most likely be developed but will require an active and sustained development program. Current fusion reactor concepts were examined as to their suitability for transmuting the separated fission wastes. It was concluded that the long-lived fission products were most amenable to transmutation. The medium-lived fission products, Cs-137 and Sr-90, require higher neutron fluxes than are available in the most developed fusion reactor concepts. Concepts which are less developed may eventually be adaptable as transmuters of these fission products

  10. Research in the field of neutronics and nuclear data for fusion

    International Nuclear Information System (INIS)

    Batistoni, P.

    2001-01-01

    A reliable and validated nuclear database is required for the design of a fusion reactor. Neutrons produced by the fusion reactions between deuterium and tritium have a very peaked energy spectrum at 14 MeV, requiring a substantial extrapolation with respect to the database made available from fission studies. The correct evaluation of shielding properties, damage, nuclear heating and of tritium breeding performance in the blanket surrounding the reaction chamber is crucial to the correct reactor design. Moreover, the attractiveness of fusion relies in the low activation of the reactor components and in the minimal production of long-term radioactive waste that is pursued with development of low activation materials. Beside the materials development, Europe is carrying out a co-ordinated program for the development of adequate nuclear database and numerical tools, directed to evaluations, processing, application, and benchmarking of cross sections including uncertainty information. Experimental validation of data and of the relative uncertainties is also pursued, both on material samples and on more design-oriented experiments. A general view of the research work in the field of neutronics and nuclear data for fusion will be given in the presentation, with emphasis to the experimental validation activity.(author)

  11. Fusion-neutron damage in superconductors and magnet stabilizers

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; Guinan, M.W.; Kinney, J.H.

    1981-01-01

    Two NbTi and two Cu wires were irradiated with 14.8 MeV neutrons at 4.2 K to fluences of 6 to 8 x 10 20 n/m 2 , using RTNS-II. Electrical resistances of Cu were monitored during irradiation. Magnetoresistances were measured in fields up to 12.4 T before and after irradiation and after isochronal annealing up to 273 K. Critical currents of NbTi were measured after irradiation, in feilds up to 10 T. The initial rate of increase of resistivity of the Cu was found to be 2.23 x 10 -31 (Ω-m)/(n/m 2 ). This rate could be predicted from fission reactor irradiations using damage energy scaling. The maximum observed change in the NbTi critical was a decrease of 3% at 4 T. At 6, 8, and 10 T there were no significant changes

  12. Research and development of a compact fusion neutron source for humanitarian landmine detection

    International Nuclear Information System (INIS)

    Yoshikawa, K.; Masuda, K.; Yamamoto, Y.; Takamatsu, T.; Toku, H.; Nagasaki, K.; Hotta, E.; Yamauchi, K.; Ohnishi, M.; Osawa, H.

    2005-01-01

    Research and development of the advanced anti-personnel landmine detection system by using a compact discharge-type D-D fusion neutron source called IECF (Inertial-Electrostatic Confinement Fusion) are described. Landmines are to be identified through increased backscattering of neutrons by the hydrogen atoms, and specific-energy capture γ-ray emission by hydrogen and nitrogen atoms with thermalized neutrons in the landmine explosives. For this purpose, improvements of the IECF device were studied for drastic enhancement of neutron production rates of more than 10 8 n/s in pulsed operation including R and D of robust power sources, as well as analyses of envisaged detection system with multi-sensors in parallel in order to show promising and practical features of this detection system for humanitarian landmine detection, particularly, in the aridic, or dry Afghanistan deserted area, where the soil moisture remains between 3-8%, which eventually enables effectively detection of hydrogen anomaly inherent in the landmine explosives. In this paper, improvements of the IECF are focused to be described. (author)

  13. Phase-IIC experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1992-12-01

    Neutronics experiments on two types of heterogeneous blankets have been performed as the Phase-IIC experiment of JAERI/USDOE collaborative program on fusion blanket neutronics. The experimental system was used in the same geometry as the previous Phase-IIA series which was a closed geometry using neutron source enclosure of lithium carbonate. The heterogeneous blankets selected here are the beryllium edge-on and the water coolant channel assemblies. In the former the beryllium and lithium-oxide layers are piled up alternately in the front part of test blanket. In the latter, the three simulated water cooling channels are settled in the Li 2 O blanket. These are producing steep gradient of neutron flux around material boundary. The calculation accuracy and measurement method for these features is a key of interest in the experiments. The measurements were performed for tritium production rate and the other nuclear parameters as well as the previous experiments. This report describes the experimental detail and the results enough to use for the benchmark data for testing the data and method of design calculation of fusion reactors. (author)

  14. Bayesian and maximum entropy methods for fusion diagnostic measurements with compact neutron spectrometers

    International Nuclear Information System (INIS)

    Reginatto, Marcel; Zimbal, Andreas

    2008-01-01

    In applications of neutron spectrometry to fusion diagnostics, it is advantageous to use methods of data analysis which can extract information from the spectrum that is directly related to the parameters of interest that describe the plasma. We present here methods of data analysis which were developed with this goal in mind, and which were applied to spectrometric measurements made with an organic liquid scintillation detector (type NE213). In our approach, we combine Bayesian parameter estimation methods and unfolding methods based on the maximum entropy principle. This two-step method allows us to optimize the analysis of the data depending on the type of information that we want to extract from the measurements. To illustrate these methods, we analyze neutron measurements made at the PTB accelerator under controlled conditions, using accelerator-produced neutron beams. Although the methods have been chosen with a specific application in mind, they are general enough to be useful for many other types of measurements

  15. Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices.

    Science.gov (United States)

    van Eden, G G; Kvon, V; van de Sanden, M C M; Morgan, T W

    2017-08-04

    Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically reduced over time due to progressing damage accumulation. Liquid metals, however, are now gaining interest in solving the challenge of extreme heat flux hitting the reactor walls. A key advantage of liquid metals is the use of vapour shielding to reduce the plasma exhaust. Here we demonstrate that this phenomenon is oscillatory by nature. The dynamics of a Sn vapour cloud are investigated by exposing liquid Sn targets to H and He plasmas at heat fluxes greater than 5 MW m -2 . The observations indicate the presence of a dynamic equilibrium between the plasma and liquid target ruled by recombinatory processes in the plasma, leading to an approximately stable surface temperature.Vapour shielding is one of the interesting mechanisms for reducing the heat load to plasma facing components in fusion reactors. Here the authors report on the observation of a dynamic equilibrium between the plasma and the divertor liquid Sn surface leading to an overall stable surface temperature.

  16. Fusion energy research with lasers, direct drive targets, and dry wall chambers

    International Nuclear Information System (INIS)

    Sethian, J.D.; Obenschain, S.P.; Myers, M.

    2003-01-01

    We are carrying out a coordinated, focused effort to develop Laser Inertial Fusion Energy. The key components are developed in concert with one another and the science and engineering issues are addressed concurrently. Significant progress has been made in this program: We are evaluating target designs that show it could be possible to achieve the high gains (>100) needed for a practical fusion system. These have a low density CH foam that is wicked with solid DT, and over coated with a thin high-Z layer. Significant advances have been made with the two types of laser are being developed: Krypton Fluoride (KrF) gas lasers and Diode Pumped Solid State Lasers (DPPSL). Both have the potential to meet the fusion energy requirements for rep-rate, efficiency, durability and cost. This paper also presents the advances in development of chamber operating windows (target survival plus no wall erosion), final optics (aluminum at grazing incidence has high reflectivity and exceeds required laser damage threshold), target fabrication (advanced foams and high Z overcoats), and target injection (new facility for target injection and tracking studies). (author)

  17. Neutron/photon/electron shielding study for a laser-fusion facility

    International Nuclear Information System (INIS)

    Thompson, W.L.

    1977-01-01

    A Monte Carlo shielding study encompassing neutron, photon, and electron transport has been conducted for the High Energy Gas Laser Facility at the Los Alamos Scientific Laboratory. This paper describes the application of the Monte Carlo technique and several variance reduction schemes to the study. The calculations involve a geometry which is complicated in all three dimensions, a very intense 14 MeV neutron source, skyshine and deep penetrations. The facility design with 1.83 m concrete walls and a 1.52 m concrete roof is based on these calculations

  18. Review of magnetic fusion energy neutron cross section needs: neutronics viewpoint

    International Nuclear Information System (INIS)

    Dudziak, D.J.; Muir, D.W.

    1977-01-01

    In the overall context of fusion nucleonic analysis, most cross section deficiencies lie in the energy range 14 MeV and below. This review deals not only with new data requirements generated by current interest in d-Li sources but also with the needs of conventional nucleonic studies (i.e., 14-MeV source calculations). The many compilations of requirements are referenced, and the current assessment of high-priority needs is succinctly summarized. Then typical methodology and results (sensitivity and uncertainty analysis) are given for quantitative data assessments of the Tokamak Fusion Test Reactor and a fusion Experimental Power Reactor. Finally, a summary is presented of some probings into data above 14 MeV, which have potential applications for d-Li irradiation facilities, d-Be medical therapy sources, and electronuclear fuel production facilities. 2 figures, 9 tables

  19. HIGH SPATIAL RESOLUTION IMAGING OF INERTIAL FUSION TARGET PLASMAS USING BUBBLE NEUTRON DETECTORS

    International Nuclear Information System (INIS)

    FISHER, R.K.

    2003-01-01

    OAK B202 HIGH SPATIAL RESOLUTION IMAGING OF INERTIAL FUSION TARGET PLASMAS USING BUBBLE NEUTRON DETECTORS. Bubble detectors, which can detect neutrons with a spatial 5 to 30 (micro), are the most promising approach to imaging NIF target plasmas with the desired 5 (micro) spatial resolution in the target plane. Gel bubble detectors are being tested to record neutron images of ICF implosions in OMEGA experiments. By improving the noise reduction techniques used in analyzing the data taken in June 2000, we have been able to image the neutron emission from 6 · 10 13 yield DT target plasmas with a target plane spatial resolution of ∼ 140 (micro). As expected, the spatial resolution was limited by counting statistics as a result of the low neutron detection efficiency of the easy-to-use gel bubble detectors. The results have been submitted for publication and will be the subject of an invited talk at the October 2001 Meeting of the Division of Plasma Physics of the American Physical Society. To improve the counting statistics, data was taken in May 2001 using a stack of four gel detectors and integrated over a series of up to seven high-yield DT shots. Analysis of the 2001 data is still in its early stages. Gel detectors were chosen for these initial tests since the bubbles can be photographed several hours after the neutron exposure. They consist of ∼ 5000 drops (∼ 100 (micro) in diameter) of bubble detector liquid/cm 3 suspended in an inactive support gel that occupies ∼ 99% of the detector volume. Using a liquid bubble chamber detector and a light scattering system to record the bubble locations a few microseconds after the neutron exposure when the bubbles are ∼ 10 (micro) in diameter, should result in ∼ 1000 times higher neutron detection efficiency and a target plane resolution on OMEGA of ∼ 10 to 50 (micro)

  20. Integral activation experiment of fusion reactor materials with d-Li neutrons up to 55 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Moellendorff, Ulrich von [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Wada, Masayuki [Business Automation Co., Ltd., Tokyo (Japan)

    2000-03-01

    An integral activation experiment of fusion reactor materials with a deuteron-lithium neutron source was performed. Since the maximum energy of neutrons produced was 55 MeV, the experiment with associated analysis was one of the first attempts for extending the energy range beyond 20 MeV. The following keywords represent the present study: d-Li neutrons, 55 MeV, dosimetry, SAND-II, spectrum adjustment, LA-150, MCNP, McDeLi, IFMIF, fusion reactor materials, integral activation experiment, low-activation, F82H, vanadium-alloy, IEAF, ALARA, and sequential charged particle reaction. (author)

  1. 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.

  2. Micro-engineered first wall tungsten armor for high average power laser fusion energy systems

    Science.gov (United States)

    Sharafat, Shahram; Ghoniem, Nasr M.; Anderson, Michael; Williams, Brian; Blanchard, Jake; Snead, Lance; HAPL Team

    2005-12-01

    The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential

  3. Micro-engineered first wall tungsten armor for high average power laser fusion energy systems

    International Nuclear Information System (INIS)

    Sharafat, Shahram; Ghoniem, Nasr M.; Anderson, Michael; Williams, Brian; Blanchard, Jake; Snead, Lance

    2005-01-01

    The high average power laser program is developing an inertial fusion energy demonstration power reactor with a solid first wall chamber. The first wall (FW) will be subject to high energy density radiation and high doses of high energy helium implantation. Tungsten has been identified as the candidate material for a FW armor. The fundamental concern is long term thermo-mechanical survivability of the armor against the effects of high temperature pulsed operation and exfoliation due to the retention of implanted helium. Even if a solid tungsten armor coating would survive the high temperature cyclic operation with minimal failure, the high helium implantation and retention would result in unacceptable material loss rates. Micro-engineered materials, such as castellated structures, plasma sprayed nano-porous coatings and refractory foams are suggested as a first wall armor material to address these fundamental concerns. A micro-engineered FW armor would have to be designed with specific geometric features that tolerate high cyclic heating loads and recycle most of the implanted helium without any significant failure. Micro-engineered materials are briefly reviewed. In particular, plasma-sprayed nano-porous tungsten and tungsten foams are assessed for their potential to accommodate inertial fusion specific loads. Tests show that nano-porous plasma spray coatings can be manufactured with high permeability to helium gas, while retaining relatively high thermal conductivities. Tungsten foams where shown to be able to overcome thermo-mechanical loads by cell rotation and deformation. Helium implantation tests have shown, that pulsed implantation and heating releases significant levels of implanted helium. Helium implantation and release from tungsten was modeled using an expanded kinetic rate theory, to include the effects of pulsed implantations and thermal cycles. Although, significant challenges remain micro-engineered materials are shown to constitute potential

  4. Condensation of ablated first-wall materials in the cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Ladd, A.J.C.

    1985-01-01

    This report concerns problems involved in recondensing first-wall materials vaporized by x rays and pellet debris in the Cascade inertial confinement fusion reactor. It examines three proposed first-wall materials, beryllium oxide (BeO), silicon carbide (SiO), and pyrolytic graphite (C), paying particular attention to the chemical equilibrium and kinetics of the vaporized gases. The major results of this study are as follows. Ceramic materials composed of diatomic molecules, such as BeO and SiC, exist as highly dissociated species after vaporization. The low gas density precludes significant recombination during times of interest (i.e., less than 0.1 s). The dissociated species (Be, O, Si, and C) are, except for carbon, quite volatile and are thermodynamically stable as a vapor under the high temperature and low density found in Cascade. These materials are thus unsuitable as first-wall materials. This difficulty is avoided with pyrolytic graphite. Since the condensation coefficient of monatomic carbon vapor (approx. 0.5) is greater than that of the polyatomic vapor (<0.1), recondensation is assisted by the expected high degree of dissociation. The proposed 10-layer granular carbon bed is sufficient to condense all the carbon vapor before it penetrates to the BeO layer below. The effective condensation coefficient of the porous bed is about 50% greater than that of a smooth wall. An estimate of the mass flux leaving the chamber results in a condensation time for a carbon first wall of about 30 to 50 ms. An experiment to investigate condensation in a Cascade-like chamber is proposed

  5. Correlation between potential well structure and neutron production in inertial electrostatic confinement fusion

    International Nuclear Information System (INIS)

    Ohnishi, M.; Yamamoto, Y.; Yoshikawa, K.; Sato, K.H.

    1997-01-01

    The electrostatic potential well in inertial electrostatic confinement (IEC) is studied using two approaches. First, the equilibrium potential profile is obtained by solving the charge neutrality condition, i.e. n i n e , assuming the appropriate distribution functions for the ions and the electrons. The formation of a double well structure is demonstrated, with a depth depending upon the ratio between the focus radii of the electrons and the ions. The correlations between the well depth and the volume integrated neutron production due to deuterium-deuterium (DD) reactions are obtained. Second, in order to study the stability of the well, the dynamic behaviours of the potential well are calculated by performing time advancing numerical simulations on the basis of the particle in cell method. Single, double and triple wells, depending on the amount of injected ion current, are observed to be formed for ions with a monoenergetic distribution. The well in the centre of the multiwell structure is unstable and oscillates with a periods much longer than the inverse ion plasma frequency. A double well structure can be formed even for ions with a spread out energy distribution when the ion current is larger than the threshold value. The time averaged neutron production by DD fusion events is proportional to a power of the ion current involved in forming the double well structure. The results strongly suggest that the high neutron production rate should be attributed to not only the well depth but also the unstable behaviour of the potential, i.e. the intermittent peaking of the density in the centre region. A numerical simulation reveals that IEC possesses a favourable dependence of fusion reactions on the injected ion current for the application to a neutron source or a fusion reactor. (author). 9 refs, 9 figs

  6. Simultaneous usage of pinhole and penumbral apertures for imaging small scale neutron sources from inertial confinement fusion experiments.

    Science.gov (United States)

    Guler, N; Volegov, P; Danly, C R; Grim, G P; Merrill, F E; Wilde, C H

    2012-10-01

    Inertial confinement fusion experiments at the National Ignition Facility are designed to understand the basic principles of creating self-sustaining fusion reactions by laser driven compression of deuterium-tritium (DT) filled cryogenic plastic capsules. The neutron imaging diagnostic provides information on the distribution of the central fusion reaction region and the surrounding DT fuel by observing neutron images in two different energy bands for primary (13-17 MeV) and down-scattered (6-12 MeV) neutrons. From this, the final shape and size of the compressed capsule can be estimated and the symmetry of the compression can be inferred. These experiments provide small sources with high yield neutron flux. An aperture design that includes an array of pinholes and penumbral apertures has provided the opportunity to image the same source with two different techniques. This allows for an evaluation of these different aperture designs and reconstruction algorithms.

  7. [International Panel on 14 MeV Intense Neutron Source Based on Accelerators for Fusion Materials Study

    International Nuclear Information System (INIS)

    Thoms, K.R.; Wiffen, F.W.

    1991-01-01

    Both travelers were members of a nine-person US delegation that participated in an international workshop on accelerator-based 14 MeV neutron sources for fusion materials research hosted by the University of Tokyo. Presentations made at the workshop reviewed the technology developed by the FMIT Project, advances in accelerator technology, and proposed concepts for neutron sources. One traveler then participated in the initial meeting of the IEA Working Group on High Energy, High Flux Neutron Sources in which efforts were begun to evaluate and compare proposed neutron sources; the Fourth FFTF/MOTA Experimenters' Workshop which covered planning and coordination of the US-Japan collaboration using the FFTF reactor to irradiate fusion reactor materials; and held discussions with several JAERI personnel on the US-Japan collaboration on fusion reactor materials

  8. Comparison of calculated integral values using measured and calculated neutron spectra for fusion neutronics analyses

    International Nuclear Information System (INIS)

    Sekimoto, H.

    1987-01-01

    The kerma heat production density, tritum production density, and dose in a lithium-fluoride pile with a deuterium-tritum neutron source were calculated with a data processing code, UFO, from the pulse height distribution of a miniature NE213 neutron spectrometer, and compared with the values calculated with a Monte Carlo code, MORSE-CV. Both the UFO and MORSE-CV values agreed with the statistical error (less than 6%) of the MORSE-CV calculations, except for the outer-most point in the pile. The MORSE-CV values were slightly smaller than the UFO values for almost all cases, and this tendency increased with increasing distance from the neutron source

  9. Experimental study of neutron streaming through steel-walled annular ducts in reactor shields

    International Nuclear Information System (INIS)

    Toshimas, M.; Nobuo, S.

    1983-01-01

    For the purpose of providing experimental data to assess neutron streaming calculations, neutron flux measurements were performed along the axes of the steel-walled annular ducts set up in a water shield of the pool-type reactor JRR-4. An annular duct simulated the air gap around the main coolant pipe. Another duct simulated the streaming path around the primary circulating pump of the integrated-type marine reactor. A 90-deg bend annular duct was also studied. In a set of measurements, the distance Z between the core center and the duct axis and the annular gap width delta were taken as parameters, that is, Z = 0, 80, and 160 cm and delta = 2.2, 4.7, and 10.1 cm. The reaction rates and the fluxes measured by the activation method are given in terms of absolute magnitude within an accuracy of + or - 30%. An empirical formula is derived based on those measured data, which describes the axial distribution of the neutron flux in the steel-walled annular duct in reactor shields. It is expressed by a simple function of the axial distance in units of the square root of the line-of-sight area, S /SUB l/ . The accuracy of the formula is examined by taking into account the duct location with respect to the reactor core, the neutron energy, the steel wall thickness, and the media outside of the steel wall. The accuracy of the formula is, in general, <30% in the axial distance between 3√S /SUB l/ and 30√S /SUB l/

  10. Response of a carbon-walled proportional counter to 14 MeV neutrons

    International Nuclear Information System (INIS)

    Lewis, K.D.

    1982-01-01

    The response of a carbon-walled spherical proportional counter filled with a methane-based tissue-equivalent gas mixture at low pressure and irradiated with 14 MeV neutrons is first measured experimentally and is then calculated theoretically by using an analytical model. The model, called the CISS model, is derived from a consideration of four basic modes of interaction of charged particles generated in neutron-nucleus reactions with the spherical cavity of the detector. Since several quantities which have application in neutron dosimetry, radiation protection, and radiation biology make direct use of such spectra, it is desirable to have the ability to theoretically predict what is expected experimentally. Thus, a comparison between the two response curves is made. The discrepancy between them is investigated by considering several physical phenomena occurring within the detector wall which tend to distort the experimental response curve. In particular, the C(n,n',3α) reaction occurring in the detector wall gives rise to multiple events, originating from a single neutron interaction in the wall simultaneously strike the detector cavity, and are recorded as a single larger event in an experimental spectra. In the analytic model, the simultaneous entry of two charged particles into the cavity is scored as two separate smaller events, uncorrelated in their production. In this work, an effort is made to modify the analytic model prediction of the response curve by correcting for the multiple events which occur. Finally, the CISS model is used to compute mass stopping power corrections for this inhomogeneous detector

  11. Neutron-activation analysis of wall soils of ancient architectural monuments

    International Nuclear Information System (INIS)

    Khatamov, Sh.; Zhumamuratov, A.; Ibragimov, T.; Tillyaev, T.; Osinskaya, N.S.; Rakhmanova, T.P.; Pulatov, D.D.

    2001-01-01

    The simplified, relatively inexpensive, and productive multielemental neutron activation techniques for analysis of solid of the architectural monuments of Karakalpakstan have been elaborated. A comparison of the elemental composition of the wall soils of the ancient buildings, constructed at different historical periods, with the composition of the agricultural soils allows us to estimate the present ecological and agrogeochemical states of the agricultural soils and to trace changing the dynamics of about 30 chemical elements. (author)

  12. Plasma behavior and plasma-wall interaction in magnetic fusion divices

    International Nuclear Information System (INIS)

    Ohtsuka, Hideo

    1984-10-01

    To study the fundamental behavior of plasma in magnetic field is the main subject in the early stage of the magnetic fusion research. At the next stage, it is necessary to overcome some actual problems in order to attain reactor grade plasmas. One of them is to control impurities in the plasma. In these points of view, we carried out several experiments or theoretical analyses. Firstly, anomalous loss mechanisms in magnetic field were investigated in a toroidal multipole device JFT-1 and the role of motions of charged particles in the magnetic field was exhibited. Various measurements of plasma in the scrape-off layer were made in a divertor tokamak JFT-2a and in an ordinary tokamak JFT-2. The former study demonstrated the first successful divertor operation of the tokamak device and the latter one clarified the mechanism of arcing on the tokamak first wall. As to arcing, a new theory which describes the retrograde motion, the well known strange motion of arcs in a magnetic field, was proposed. Good agreement with the experimental results was shown. Finally, by considering a zero-dimensional sputtering model a self-consistent relation between light and metal impurities in tokamak plasmas was obtained. It was shown that the relation well describes some fundamental aspects of the plasma-wall interaction. As a conclusion, the importance of simple behavior of charged particles in magnetic fields was pointed out not only for the plasma confinement but also for the plasma-wall interaction. (author)

  13. Refractory oxides for fusion reactor first walls, the effects of the reducing environment

    International Nuclear Information System (INIS)

    Hoffman, J.G.

    1979-01-01

    Of the several applications for refractory oxides in fusion reactor systems, the most demanding is that for the first wall. Some components in proximity of the first wall (possibly waveguides or flux breakers) will also be subjected to similar environments. Many parameters affect the ultimate usability of a particular material for reactor applications: electrical resistivity and dielectric breakdown if applicable, thermal conductivity, mechanical properties, and stability with respect to neutral molecular or atomic, or ionized fuel gases. All these properties can be affected by the radiation environment present in an operating power reactor. Temperatures up to 2000K may be expected for radiatively cooled first wall liners in some proposed designs although surface temperatures are appreciably lower (approximately 1000K) in other applications. The exact nature of the chemical environment is not defined even for the most well developed design concepts, but possible environments may be hypothesized; ambient neutral molecular and atomic species, bombardment by high energy charge exchange neutral atoms, direct ionic bombardment from stray ions, and plasma dumps from failure of the confinement system. Preliminary work has begun to more adequately define the extent of the problem and suggest approaches to engineering solutions

  14. A conceptual design strategy for liquid-metal-wall inertial fusion reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade are reviewed from the perspective of formulating a conceptual design strategy for such chambers. The basis for the design strategy is set by enumerating both the attractive and unattractive features of a LMW chamber. Past concepts are then reviewed to identify conceptual design approaches and physical configurations that enhance the positive aspects and minimize the negative aspects. A detailed description of the engineering considerations is given, including such topics as the selection of a liquid metal, control of radiation damage, selection of structural material, control of tritium breeding and extraction, control of wall stress, and designing for a given rep-rate. Finally, a design strategy is formulated which accomodates the engineering constraints while minimizing the liquid-metal flow rate. (orig.)

  15. Neutron imaging development for megajoule scale inertial confinement fusion experiments{sup 1}

    Energy Technology Data Exchange (ETDEWEB)

    Grim, G P; Bradley, P A; Day, R D; Clark, D D; Fatherley, V E; Finch, J P; Garcia, F P; Jaramillo, S A; Montoya, A J; Morgan, G L; Oertel, J A; Ortiz, T A; Payton, J R; Pazuchanics, P; Schmidt, D W; Valdez, A C; Wilde, C H; Wilke, M D; Wilson, D C [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)], E-mail: gpgrim@lanl.gov

    2008-05-15

    Neutron imaging of Inertial Confinement Fusion (ICF) targets is useful for understanding the implosion conditions of deuterium and tritium filled targets at Mega-Joule/Tera-Watt scale laser facilities. The primary task for imaging ICF targets at the National Ignition Facility, Lawrence Livermore National Laboratory, Livermore CA, is to determine the asymmetry of the imploded target. The image data, along with other nuclear information, are to be used to provide insight into target drive conditions. The diagnostic goal at the National Ignition Facility is to provide neutron images with 10 {mu}m resolution and peak signal-to-background values greater than 20 for neutron yields of {approx} 10{sup 15}. To achieve this requires signal multiplexing apertures with good resolution. In this paper we present results from imaging system development efforts aimed at achieving these requirements using neutron pinholes. The data were collected using directly driven ICF targets at the Omega Laser, University of Rochester, Rochester, NY., and include images collected from a 3 x 3 array of 15.5 {mu}m pinholes. Combined images have peak signal-to-background values greater than 30 at neutron yields of {approx} 10{sup 13}.

  16. Neutronics analysis of minor actinides transmutation in a fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Yang, Chao; Cao, Liangzhi; Wu, Hongchun; Zheng, Youqi; Zu, Tiejun

    2013-01-01

    Highlights: • A fusion fission hybrid system for MA transmutation is proposed. • The analysis of neutronics effects on the transmutation is performed. • The transmutation rate of MA reaches 86.5% by 25 times of recycling. -- Abstract: The minor actinides (MAs) transmutation in a fusion-driven subcritical system is analyzed in this paper. The subcritical reactor is driven by a tokamak D-T fusion device with relatively easily achieved plasma parameters and tokamak technologies. The MAs discharged from the light water reactor (LWR) are loaded in transmutation zone. Sodium is used as the coolant. The mass percentage of the reprocessed plutonium (Pu) in the fuel is raised from 0 to 48% and stepped by 12% to determine its effect on the MAs transmutation. The lesser the Pu is loaded, the larger the MAs transmutation rate is, but the smaller the energy multiplication factor is. The neutronics analysis of two loading patterns is performed and compared. The loading pattern where the mass percentage of Pu in two regions is 15% and 32.9% respectively is conducive to the improvement of the transmutation fraction within the limits of burn-up. The final transmutation fraction of MAs can reach 17.8% after five years of irradiation. The multiple recycling is investigated. The transmutation fraction of MAs can reach about 61.8% after six times of recycling, and goes up to about 86.5% after 25

  17. Investigation of lanthanides as neutron multipliers for hybrid and fusion reactor blankets

    International Nuclear Information System (INIS)

    Sahin, Sumer

    1982-01-01

    The neutronic performance of three lanthanides ( 149 Sm, europium, and gadolinium) as neutron multiplier for the blanket of a fusion-fission (hybrid) and a pure fusion reactor has been evaluated and compared with that of beryllium and lead. During the calculations, the fission zone is made up of UO 2 rods from the LOTUS experimental hybrid facility now under construction at the Nuclear Engineering Laboratory of the Swiss Federal Institute of Technology in Lausanne. In fusion blanket the fuel zone is replaced by pure lithium. The calculations were performed for two different boundary conditions for the left boundary: (a) reflecting, representative of a typical confinement geometry, and (b) vacuum, which represents a typical blanket experiment in plane geometry. For a vacuum left boundary, threshold reactions are reduced by a factor of about 2 while 1/v-type reactions are decreased by a factor of between 5 and 10, as a consequence of the softer spectrum produced by a reflecting left boundary. In general, the results, notably tritium breeding and energy multiplication, are comparable for the lanthanide multipliers and for beryllium and lead if the left boundary is a vacuum. The use of 149 Sm is slightly less effective than europium or gadolinium and all of the lanthanides perform better for a vacuum left boundary than for the reflecting case. The analyses presented here also illustrate the importance of potential spectral shifts that can occur as the result of experimental exigencies

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  19. Summary report for IAEA CRP on lifetime prediction for the first wall of a fusion machine (JAERI contribution)

    International Nuclear Information System (INIS)

    Suzuki, Satoshi; Araki, Masanori; Akiba, Masato

    1993-03-01

    IAEA Coordinated Research Program (CRP) on 'Lifetime Prediction for the First Wall of a Fusion Machine' was started in 1989. Five participants, Joint Research Centre (JRC-Ispra), The NET team, Kernforschungszentrum Karlsruhe (KfK), Russian Research Center and Japan Atomic Energy Research Institute, contributed in this activity. The purpose of the CRP is to evaluate the thermal fatigue behavior of the first wall of a next generation fusion machine by means of numerical methods and also to contribute the design activities for ITER (International Thermonuclear Experimental Reactor). Thermal fatigue experiments of a first wall mock-up which were carried out in JRC-Ispra were selected as a first benchmark exercise model. All participants performed finite element analyses with various analytical codes to predict the lifetime of the simulated first wall. The first benchmark exercise has successfully been finished in 1992. This report summarizes a JAERI's contribution for this first benchmark exercise. (author)

  20. Liquid Wall Options for Tritium-Lean Fast Ignition Inertial Fusion Energy Power Plants

    International Nuclear Information System (INIS)

    Reyes, S.; Schmitt, R.C.; Latkowski, J.F.; Durbin, S.G.' Sanz, J.

    2002-01-01

    In an inertial fusion energy (FE) thick-liquid chamber design such as HYLEE-II, a molten-salt is used to attenuate neutrons and protect the chamber structures from radiation damage. In the case of a fast ignition inertial fusion system, advanced targets have been proposed that may be self-sufficient in terms of tritium breeding (i.e., the amount of tritium bred in target exceeds the amount burned). This aspect allows for greater freedom when selecting a liquid for the protective blanket, given that lithium-bearing compounds are no longer required. The present work assesses the characteristics of many single, binary, and ternary molten-salts using the NIST Properties of Molten Salts Database. As an initial screening, salts were evaluated for their safety and environmental (S and E) characteristics, which included an assessment of waste disposal rating, contact dose, and radioactive afterheat. Salts that passed the S and E criteria were then evaluated for required pumping power. The pumping power was calculated using three components: velocity head losses, frictional losses, and lifting power. The results of the assessment are used to identify those molten-salts that are suitable for potential liquid-chamber fast-ignition IFE concepts, from both the S and E and pumping power perspective. Recommendations for further analysis are also made

  1. Generation of thermonuclear fusion neutrons by means of a pure explosion. Part 2. Experimental results

    International Nuclear Information System (INIS)

    Derentowicz, H.; Kaliski, S.; Wolski, J.; Ziolkowski, Z.

    1977-01-01

    This paper presents the experimental results of the generation of a thermonuclear fusion neutrons by means of explosion. The experimental set is based on a quasi-spherical experiment in which a polyethylene layer is shot into a conic region hollowed out in a golden target and filled with deuterium gas. The speeding-up system is based on shooting the conic liner onto the surface of the Cu cone in which the Mach wave is generated and propagates along the cone axis leading to an implosion velocity of the polyethylene layer of the order of (4 - 5).10 6 cm/s. This affords a 10 3 -multiple compression of the D 2 gas (p 0 approximately 1.2 atm) and a neutron emission of the order of 3.10 7 from a mass of about 10 -7 g. This result is in full agreement with theoretical estimates. This is the first published and documented experiment in which a neutron stream of thermonuclear fusion was obtained by means of a pure explosion. (author)

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

    International Nuclear Information System (INIS)

    Beller, D.E.

    1986-01-01

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

  3. Second international comparison on measuring techniques of tritium production rate for fusion neutronics experiments (ICMT-2)

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Maekawa, Hiroshi

    1993-02-01

    An second international comparison on measuring techniques of tritium production rates for fusion neutronics experiments (ICMT-2) has been performed. The purpose is to evaluate the measurement accuracy of tritium production rates in the current measurement techniques. Two 14 MeV neutron source facilities, FNS at JAERI-Japan and LOTUS at EPFL-Switzerland, were used for this purpose. Nine groups out of seven countries participated in this program. A fusion simulated blanket assembly of simple-geometry was served as the test bed at each facility, in which Li-containing samples from the participants were irradiated in an uniform neutron field. The tritium production rates were determined by the participants using their own ways by using the liquid scintillation counting method. Tritiated water sample with unknown but the same concentration was also distributed and its concentration was measured to make a common reference. The standard deviation of measured tritium production rates among participants was about 10 % for both FNS and LOTUS irradiation levels: 4x10 -13 T-atoms/Li-atom and 1.6x10 -12 T-atoms/Li-atom at a sample, respectively. This standard deviation exceeds the expected deviation of 5 % in this program. It is presumed that the deviation of 10 % is caused mainly by the systematic and unknown errors in a process of tritium extraction from the irradiated samples depending on each organization. (author)

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

    CERN Document Server

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

    2000-01-01

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

  5. Stability of the lithium waterfall first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion (ICF) reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived which predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  6. Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  7. Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  8. A data fusion approach for progressive damage quantification in reinforced concrete masonry walls

    International Nuclear Information System (INIS)

    Vanniamparambil, Prashanth Abraham; Carmi, Rami; Kontsos, Antonios; Bolhassani, Mohammad; Khan, Fuad; Bartoli, Ivan; Moon, Franklin L; Hamid, Ahmad

    2014-01-01

    This paper presents a data fusion approach based on digital image correlation (DIC) and acoustic emission (AE) to detect, monitor and quantify progressive damage development in reinforced concrete masonry walls (CMW) with varying types of reinforcements. CMW were tested to evaluate their structural behavior under cyclic loading. The combination of DIC with AE provided a framework for the cross-correlation of full field strain maps on the surface of CMW with volume-inspecting acoustic activity. AE allowed in situ monitoring of damage progression which was correlated with the DIC through quantification of strain concentrations and by tracking crack evolution, visually verified. The presented results further demonstrate the relationships between the onset and development of cracking with changes in energy dissipation at each loading cycle, measured principal strains and computed AE energy, providing a promising paradigm for structural health monitoring applications on full-scale concrete masonry buildings. (paper)

  9. Plasma Surface Interactions Common to Advanced Fusion Wall Materials and EUV Lithography - Lithium and Tin

    Science.gov (United States)

    Ruzic, D. N.; Alman, D. A.; Jurczyk, B. E.; Stubbers, R.; Coventry, M. D.; Neumann, M. J.; Olczak, W.; Qiu, H.

    2004-09-01

    Advanced plasma facing components (PFCs) are needed to protect walls in future high power fusion devices. In the semiconductor industry, extreme ultraviolet (EUV) sources are needed for next generation lithography. Lithium and tin are candidate materials in both areas, with liquid Li and Sn plasma material interactions being critical. The Plasma Material Interaction Group at the University of Illinois is leveraging liquid metal experimental and computational facilities to benefit both fields. The Ion surface InterAction eXperiment (IIAX) has measured liquid Li and Sn sputtering, showing an enhancement in erosion with temperature for light ion bombardment. Surface Cleaning of Optics by Plasma Exposure (SCOPE) measures erosion and damage of EUV mirror samples, and tests cleaning recipes with a helicon plasma. The Flowing LIquid surface Retention Experiment (FLIRE) measures the He and H retention in flowing liquid metals, with retention coefficients varying between 0.001 at 500 eV to 0.01 at 4000 eV.

  10. Fabrication of thin-wall, freestanding inertial confinement fusion targets by chemical vapor deposition

    International Nuclear Information System (INIS)

    Carroll, D.W.; McCreary, W.J.

    1982-01-01

    To meet the requirements for plasma physics experiments in the inertial confinement fusion (ICF) program, chemical vapor deposition (CVD) in fluid beds was used to fabricate freestanding tungsten spheres and cylinders with wall thicknesses less than 5.0 μm. Molybdenum and molybdenum alloy (TZM) mandrels of the desired geometry were suspended in a carrier bed of dense microspheres contained in an induction-heated fluid-bed reactor. The mandrels were free to float randomly through the bed, and using the reaction WF 6 +3H 2 →/sub /KW +6HF, very fine-grained tungsten was deposited onto the surface at a rate and in a grain size determined by temperature, gas flow rate, system pressure, and duration of the reaction. After coating, a portion of each mandrel was exposed by hole drilling or grinding. The mandrel was then removed by acid leaching, leaving a freestanding tungsten shape. Experimental procedures, mandrel preparation, and results obtained are discussed

  11. Neutron irradiation effects in fusion or spallation structural materials: Some recent insights related to neutron spectra

    International Nuclear Information System (INIS)

    Garner, F.A.; Greenwood, L.R.

    1998-01-01

    A review is presented of recent insights on the role of transmutation in the development of radiation-induced changes in dimension or radiation-induced changes in physical or mechanical properties. It is shown that, in some materials and some neutron spectra, transmutation can significantly affect or even dominate a given property change process. When the process under study is also sensitive to displacement rate, and especially if it involves radiation-induced segregation and precipitation, it becomes much more difficult to separate the transmutation and displacement rate dependencies. This complicates the application of data derived from 'surrogate' spectra to predictions in other flux-spectra environments. It is also shown in this paper that one must be sensitive to the impact of previously -ignored 'small' variations in neutron spectra within a given reactor. In some materials these small variations have major consequences. (author)

  12. Radiological consequences of a bounding event sequence of Advanced Fusion Neutron Source (A-FNS)

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Makoto M., E-mail: nakamura.makoto@qst.go.jp; Ochiai, Kentaro

    2017-05-15

    Advanced Fusion Neutron Source (A-FNS) is an accelerator-based neutron source utilizing Li(d,xn) nuclear stripping reactions to simulate D-T fusion neutrons for testing and qualifying structural and functional materials of fusion reactor components, which is to be constructed at the Rokkasho site of National Institutes for Quantum and Radiological Science and Technology, Japan, in the near future. The purpose of the study reported here is to demonstrate the ultimate safety margins of A-FNS in the worst case of release of radioactive materials outside the A-FNS confinement system. For this purpose, we analyzed a ‘bounding event’ postulated in A-FNS. The postulated event sequence consists of fire of the purification system of the liquid Li loop during the maintenance, of mobilization of the tritium and {sup 7}Be, which are the impurities of the loop, and of the entire loss of confinement of the radioactive materials. We have calculated the early doses to the public due to the release of the tritium and {sup 7}Be source terms to the environment. The UFOTRI/COSYMA simulations have been performed considering the site boundary of 500 m away from the facility. The obtained results indicate that the early dose is below the level that requires the emergent public evacuation. Such results demonstrate that the A-FNS complies with the defined safety objective against its radiation hazard. The simulation results suggest that the inherent, ultimate safety characteristic found by this study may assist a licensing process for installation of A-FNS.

  13. Development of fatigue life criteria for experimental fusion reactor first-wall structures

    International Nuclear Information System (INIS)

    Nickell, R.E.; Esztegar, E.P.

    1980-01-01

    An approach to the rational design of fusion reactor first-wall structures against fatigue crack growth is proposed. The approach is motivated by microstructural observations of fatigue crack growth enhancement in uniruniradiated materials due to volumetric damage ahead of a propagating crack. Examples are cited that illustrate the effect of mean stress on void nucleation and coalescence, which represent the dominant form of volumetric damage at low temperature, and of grain boundary sliding and creep cavitation, which are the dominant volumetric damage mechanisms at high temperature. The analogy is then drawn between these forms of fatigue crack growth enhancement and those promoted by irradiation exposure in the fusion reactor environment, such as helium embrittlement and atomic displacement. An enhanced strain range is suggested as a macroscopic measure of the reduction in fatigue life due to the higher fatigue crack growth rates. The enhanced strain range permits a separation of volumetric and cyclic effects, and assists in the assignment of rational design factors to each effect. A series of experiments are outlined which should provide the numerical values of the parameters for the enhanced strain range. (orig.)

  14. Spectra of neutrons and fusion charged products produced in a dense laser plasma

    International Nuclear Information System (INIS)

    Burtsev, V.A.; Dyatlov, V.D.; Krzhizhanovskij, R.E.; Levkovskij, A.A.

    1977-01-01

    The possibility of laser-produced plasma diagnostics has been investigated by measuring spectra of neutrons and alpha particles produced in the T(d,n) 4 He reaction. Using the Monte Carlo method the spectra have been calculated for nine states of the deuterium-tritium plasma with the temperature of 1;5 and 10 keV and the density of 0.2; 1 and 10 g/cm 3 respectively. The initial radius of the target was assumed to be 0.01 cm at the density of 0.2 g/cm 3 . It is shown that the neutron and alpha spectra can serve as plasma diagnostics parameters in laser fusion

  15. Design and construction of a fast (neutron) spectrometer for magnetic confinement fusion studies

    International Nuclear Information System (INIS)

    Lagattu, A.

    1980-09-01

    Neutron emission in conjunction with the 'Tokamak' devices studied in order to carry out controlled fusion is done at present on a short time scale (a few hundred milliseconds). Obtaining spectrometric information on this emission made it necessary to design sufficiently fast equipment to adapt to this situation. The basis of the equipment selected is conventional: NE213 liquid scintillator in conjunction with a fast photomultiplier. A commercialized fast neutron gamma discriminator (approximately 210 5 S -1 ) was used. The data acquisition and processing system was specially made for this occasion. This assembly, so designed and constructed, was tested on a particle accelerator and on the TFR Tokamak. In the best conditions a spectrum of 60 milliseconds was produced. Hence several spectra can be recorded during an experiment. The rejection of gamma radiations by the discriminator exceeds 99% in such conditions [fr

  16. Tritium breeding experiments in a fusion blanket assembly using a low-intensity neutron generator

    International Nuclear Information System (INIS)

    Dalton, A.W.; Woodley, H.J.; McGregor, B.J.

    1987-01-01

    Experiments have been carried out to determine the accuracy with which tritium production rates (TPRs) can be measured in a fusion blanket assembly of non-spherical geometry by a non-central low intensity D-T neutron source (2x10 10 neutrons per second). The tritium production was determined for samples of lithium carbonate containing high enrichments of 6 Li(96%) and 7 Li(99.9%). The measured data were used to check the accuracy with which the TPRs could be numerically predicted using current nuclear data and calculational methods. The numerical predictions from tritium production from the 7 Li samples agreed within the experimental errors of the measurements, but 6 Li measurements which differ by more than 20 per cent from the predicted values were observed in the lower half of the assembly

  17. Neutron and Proton Diffusion in Fusion Reactions for the Synthesis of Superheavy Nuclei

    International Nuclear Information System (INIS)

    Ming-Hui, Huang; Zai-Guo, Gan; Zhao-Qing, Feng; Xiao-Hong, Zhou; Jun-Qing, Li

    2008-01-01

    The restriction of the one dimensional (1D) master equation (ME) with the mass number of the projectile-like fragment as a variable is studied, and a two-dimensional (2D) master equation with the neutron and proton numbers as independent variables is set up, and solved numerically. Our study showed that the 2D ME can describe the fusion process well in all projectile-target combinations. Therefore the possible channels to synthesize super-heavy nuclei can be studied correctly in wider possibilities. The available condition for employing 1D ME is pointed out

  18. Neutron spectra and level density parameters from 16O + 12C fusion reaction

    International Nuclear Information System (INIS)

    Kasagi, J.; Remington, B.; Galonsky, A.; Haas, F.; Racca, R.; Prosser, F.W.

    1985-01-01

    Residues following 16 O + 12 C fusion were identified by their characteristic γ-rays. For several transitions in 23 Mg, 25 Mg, and 26 Al coincident neutron spectra were measured at six angles. Through use of the evaporation code CASCADE, comparisons were made of these spectra with predictions of the statistical model at five 16 O projectile energies between 43.2 and 56.0 MeV. The results require an excitation energy dependence for the effective radius parameter r 0 which determines the spin cutoff factor

  19. Liquid metal coolants for fusion-fission hybrid system: A neutronic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Renato V.A.; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., E-mail: claubia@nuclear.ufmg.br [Universidade de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany P. [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Based on a work already published by the UFMG Nuclear Engineering Department, it was suggested to use different coolant materials in a fusion-fission system after a fuel burnup simulation, including that one used in reference work. The goal is to compare the neutron parameters, such as the effect multiplication factor and actinide amounts in transmutation layer, for each used coolant and find the best(s) coolant material(s) to be applied in the considered system. Results indicate that the lead and lead-bismuth coolant are the most suitable choices to be applied to cool the system. (author)

  20. High resolution neutron diffraction crystallographic investigation of Oxide Dispersion Strengthened steels of interest for fusion technology

    International Nuclear Information System (INIS)

    Coppola, R.; Rodriguez-Carvajal, J.; Wang, M.; Zhang, G.; Zhou, Z.

    2014-01-01

    High resolution neutron diffraction measurements have been carried out to characterize the crystallographic phases present in different Oxide Dispersion Strengthened (ODS) steels of interest for fusion technology. The different lattice structures, Im3m for the ferritic ODS and Fm3m for the austenitic ODS, are resolved showing line anisotropy effects possibly correlated with differences in dislocation densities and texture. Many contributions from minority phases are detected well above the background noise; none of the expected crystallographic phases, such as M 23 C 6 and including Y 2 O 3 , fits them, but the TiN phase is identified in accordance with results of other microstructural techniques

  1. INDRA: a program system for calculating the neutronics and photonics characteristics of a fusion reactor blanket

    International Nuclear Information System (INIS)

    Perry, R.T.; Gorenflo, H.; Daenner, W.

    1976-01-01

    INDRA is a program system for calculating the neutronics and photonics characteristics of fusion reactor blankets. It incorporates a total of 19 different codes and 5 large data libraries. 10 of the codes are available from the code distribution organizations. Some of them, however, have been slightly modified in order to permit a convenient transfer of information from one program module to the next. The remaining 9 programs have been prepared by the authors to complete the system with respect to flexibility and to facilitate the handling of the results. (orig./WBU) [de

  2. Peaceful fusion

    Energy Technology Data Exchange (ETDEWEB)

    Englert, Matthias [IANUS, TU Darmstadt (Germany)

    2014-07-01

    Like other intense neutron sources fusion reactors have in principle a potential to be used for military purposes. Although the use of fissile material is usually not considered when thinking of fusion reactors (except in fusion-fission hybrid concepts) quantitative estimates about the possible production potential of future commercial fusion reactor concepts show that significant amounts of weapon grade fissile materials could be produced even with very limited amounts of source materials. In this talk detailed burnup calculations with VESTA and MCMATH using an MCNP model of the PPCS-A will be presented. We compare different irradiation positions and the isotopic vectors of the plutonium bred in different blankets of the reactor wall with the liquid lead-lithium alloy replaced by uranium. The technical, regulatory and policy challenges to manage the proliferation risks of fusion power will be addressed as well. Some of these challenges would benefit if addressed at an early stage of the research and development process. Hence, research on fusion reactor safeguards should start as early as possible and accompany the current research on experimental fusion reactors.

  3. Extensive neutronic sensitivity-uncertainty analysis of a fusion reactor shielding blanket

    International Nuclear Information System (INIS)

    Hogenbirk, A.

    1994-01-01

    In this paper the results are presented of an extensive neutronic sensitivity-uncertainty study performed for the design of a shielding blanket for a next-step fusion reactor, such as ITER. A code system was used, which was developed at ECN Petten. The uncertainty in an important response parameter, the neutron heating in the inboard superconducting coils, was evaluated. Neutron transport calculations in the 100 neutron group GAM-II structure were performed using the code ANISN. For the sensitivity and uncertainty calculations the code SUSD was used. Uncertainties due to cross-section uncertainties were taken into account as well as uncertainties due to uncertainties in energy and angular distributions of scattered neutrons (SED and SAD uncertainties, respectively). The subject of direct-term uncertainties (i.e. uncertainties due to uncertainties in the kerma factors of the superconducting coils) is briefly touched upon. It is shown that SAD uncertainties, which have been largely neglected until now, contribute significantly to the total uncertainty. Moreover, the contribution of direct-term uncertainties may be large. The total uncertainty in the neutron heating, only due to Fe cross-sections, amounts to approximately 25%, which is rather large. However, uncertainty data are scarce and the data may very well be conservative. It is shown in this paper that with the code system used, sensitivity and uncertainty calculations can be performed in a straightforward way. Therefore, it is suggested that emphasis is now put on the generation of realistic, reliable covariance data for cross-sections as well as for angular and energy distributions. ((orig.))

  4. Integral tests of coupled multigroup neutron and gamma cross sections with fission and fusion sources

    International Nuclear Information System (INIS)

    Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.

    1978-01-01

    Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)

  5. Superresolution of a compact neutron spectrometer at energies relevant for fusion diagnostics

    International Nuclear Information System (INIS)

    Reginatto, M.; Zimbal, A.

    2011-01-01

    The ability to achieve resolution that is better than the instrument resolution (i.e., superresolution) is well known in optics, where it has been extensively studied. Unfortunately, there are only a handful of theoretical studies concerning superresolution of particle spectrometers, even though experimentalists are familiar with the enhancement of resolution that is achievable when appropriate methods of data analysis are used, such as maximum entropy and Bayesian methods. Knowledge of the superresolution factor is in many cases important. For example, in applications of neutron spectrometry to fusion diagnostics, the temperature of a burning plasma is an important physical parameter which may be inferred from the width of the peak of the neutron energy spectrum, and the ability to determine this width depends on the superresolution factor. Kosarev has derived an absolute limit for resolution enhancement using arguments based on a well known theorem of Shannon. Most calculations of superresolution factors in the literature, however, are based on the assumption of Gaussian, translationally invariant response functions and therefore not directly applicable to neutron spectrometers which typically have response functions not satisfying these requirements. In this work, we develop a procedure that allows us to overcome these difficulties and we derive estimates of superresolution for liquid scintillator spectrometers of a type commonly used for neutron measurements. Theoretical superresolution factors are compared to experimental results.

  6. A compact neutron counter telescope with thick radiator (COTETRA) for fusion experiment

    International Nuclear Information System (INIS)

    Osakabe, M.; Itoh, S.; Gotoh, Y.; Sasao, M.; Fujita, J.

    1993-01-01

    A new type of neutron spectrometer, based on recoil proton measurement, has been developed for diagnosing a DT fusion plasma. This spectrometer has such advantages as: 1. direct measurement of the neutron energy without the unfolding procedure, 2. relatively high detection efficiency for 14 MeV neutrons, 3. a wide dynamic range of counting rate, and 4. perfect n-γ discrimination. To examine the performance of this spectrometer, we developed a Monte Carlo simulation code. It predicts that we may achieve energy resolution up to 3 % with a detection efficiency of 10 -5 [count cm 2 /n] if we could successfully adjust the condition of the spectrometer. A prototype spectrometer was constructed and was compared with the Monte Carlo prediction. The energy resolution of 5.3 ± 0.7 % for 14 MeV neutron was obtained for the prototype spectrometer and the calculation agrees with the experimental results within its margin of error if we take into account the intrinsic energy resolution of the detector that is used in the prototype. (author)

  7. On the absolute calibration of a DT fusion neutron yield diagnostic

    Directory of Open Access Journals (Sweden)

    Ruiz C.L.

    2013-11-01

    Full Text Available Recent advances in Inertial Confinement Fusion (ICF experiments at Lawrence Livermore National Laboratory's National Ignition Facility (NIF have underscored the need for accurate total yield measurements of DT neutrons because yield measurements provide a measure of the predicted performance of the experiments. Future gas-puff DT experiments at Sandia National Laboratory's Z facility will also require similar measurements. For ICF DT experiments, the standard technique for measuring the neutron (14.1 MeV yield, counts the activity (counts/minute induced in irradiated copper samples. This activity occurs by the 63Cu(n,2n62Cu reaction where 62Cu decays by positrons (β+ with a half-life of 9.67 minutes. The calibrations discussed here employ the associated-particle method (APM, where the α (4He particles from the T(d,n4He reaction are measured to infer neutron fluxes on a copper sample. The flux induces 62Cu activity, measured in a coincidence counting system. The method leads to a relationship between a DT neutron yield and copper activity known as the F-factor. The goal in future experiments is to apply this calibration to measure the yield at NIF with a combined uncertainty approaching 5%.

  8. FENDL-3 benchmark test with neutronics experiments related to fusion in Japan

    International Nuclear Information System (INIS)

    Konno, Chikara; Ohta, Masayuki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi

    2014-01-01

    Highlights: •We have benchmarked FENDL-3.0 with integral experiments with DT neutron sources in Japan. •The FENDL-3.0 is as accurate as FENDL-2.1 and JENDL-4.0 or more. •Some data in FENDL-3.0 may have some problems. -- Abstract: The IAEA supports and promotes the gathering of the best data from evaluated nuclear data libraries for each nucleus involved in fusion reactor applications and compiles these data as FENDL. In 2012, the IAEA released a major update to FENDL, FENDL-3.0, which extends the neutron energy range from 20 MeV to greater than 60 MeV for 180 nuclei. We have benchmarked FENDL-3.0 versus in situ and TOF experiments using the DT neutron source at FNS at the JAEA and TOF experiments using the DT neutron source at OKTAVIAN at Osaka University in Japan. The Monte Carlo code MCNP-5 and the ACE file of FENDL-3.0 supplied from the IAEA were used for the calculations. The results were compared with measured ones and those obtained using the previous version, FENDL-2.1, and the latest version, JENDL-4.0. It is concluded that FENDL-3.0 is as accurate as or more so than FENDL-2.1 and JENDL-4.0, although some data in FENDL-3.0 may be problematic

  9. Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices

    NARCIS (Netherlands)

    Philipps, V.; Malaquias, A.; Hakola, A.; Karhunen, J.; Maddaluno, G.; Almaviva, S.; Caneve, L.; Colao, F.; Fortuna, E.; Gasior, P.; Kubkowska, M.; Czarnecka, A.; Laan, M.; Lissovski, A.; Paris, P.; van der Meiden, H. J.; Petersson, P.; Rubel, M.; Huber, A.; Zlobinski, M.; Schweer, B.; Gierse, N.; Xiao, Q.; Sergienko, G.

    2013-01-01

    Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the

  10. Phase IIA and IIB experiments of JAERI/U.S.DOE collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1989-12-01

    Phase IIA and IIB experiments on fusion blanket neutronics has been performed on a basis of JAERI/USDOE collaborative program. In the Phase II experimental series, a D-T neutron source and a test blanket were contained by a lithium-carbonate enclosure to adjust the incident neutron spectrum to the test blanket so as to simulate that of a fusion reactor. First two series of the Phase II, IIA and IIB, focused especially on influences of beryllium configurations for neutron multiplying zone to neutronic parameters. Measured parameters were tritium production rate using Li-glass and NE213 scintillators, and Li-metal foil and Lithium-oxide block with liquid scintillation technique; neutron spectrum using NE213 scintillator and proton recoil proportional counter; reaction rate using foil activation technique. These parameters were compared among six different beryllium configurations of the experimental system. Consistency between different techniques for each measured parameter was also tested among different experimental systems and confirmed to be within experimental errors. This report describes, in detail, experimental conditions, assemblies, equipments and neutron source in Part I. The part II compiles all information required for a calculational analysis of this experiment, e.g., dimensions of the target room, target assembly, experimental assembly, their material densities and numerical data of experimental results. This compilation provides benchmark data to test calculation models and computing code systems used for a nuclear design of a fusion reactor. (author)

  11. Rationalization and utilization of double-wall vacuum vessel for tokamak fusion facility

    International Nuclear Information System (INIS)

    Nakahira, Masataka

    2005-09-01

    Vacuum Vessel (VV) of ITER is difficult to apply a non-destructive in-service inspection (ISI) and then new safety concept is needed. Present fabrication standards are not applicable to the VV, because the access is limited to the backside of closure weld of double wall. Fabrication tolerance of VV is ± 5mm even the structure is huge as high as 10m. This accuracy requires a rational method on the estimation of welding deformation. In this report, an inherent safety feature of the tokamak is proved closing up a special characteristic of termination of fusion reaction due to tiny water leak. A rational concept not to require ISI without sacrificing safety is shown based on this result. A partial penetration T-welded joint is proposed to establish a rational fabrication method of double wall. Strength and susceptibility to crevice corrosion is evaluated for this joint and feasibility is confirmed. A rational method of estimation of welding deformation for large and complex structure is proposed and the efficiency is shown by comparing analysis and experimental results of full-scale test. (author)

  12. Development of a low activation concrete shielding wall by multi-layered structure for a fusion reactor

    International Nuclear Information System (INIS)

    Sato, Satoshi; Maegawa, Toshio; Yoshimatsu, Kenji; Sato, Koichi; Nonaka, Akira; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara

    2011-01-01

    A multi-layered concrete structure has been developed to reduce induced activity in the shielding for neutron generating facilities such as a fusion reactor. The multi-layered concrete structure is composed of: (1) an inner low activation concrete, (2) a boron-doped low activation concrete as the second layer, and (3) ordinary concrete as the outer layer of the neutron shield. With the multi-layered concrete structure the volume of boron is drastically decreased compared to a monolithic boron-doped concrete. A 14 MeV neutron shielding experiment with multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activity in the mockups were measured. This demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activity.

  13. Research on anisotropy of fusion-produced protons and neutrons emission from high-current plasma-focus discharges

    Energy Technology Data Exchange (ETDEWEB)

    Malinowski, K., E-mail: karol.malinowski@ncbj.gov.pl; Sadowski, M. J.; Szydlowski, A. [National Centre for Nuclear Research (NCBJ), 05-400 Otwock (Poland); Institute of Plasma Physics and Laser Microfusion (IFPiLM), 01-497 Warsaw (Poland); Skladnik-Sadowska, E.; Czaus, K.; Kwiatkowski, R.; Zaloga, D. [National Centre for Nuclear Research (NCBJ), 05-400 Otwock (Poland); Paduch, M.; Zielinska, E. [Institute of Plasma Physics and Laser Microfusion (IFPiLM), 01-497 Warsaw (Poland)

    2015-01-15

    The paper concerns fast protons and neutrons from D-D fusion reactions in a Plasma-Focus-1000U facility. Measurements were performed with nuclear-track detectors arranged in “sandwiches” of an Al-foil and two PM-355 detectors separated by a polyethylene-plate. The Al-foil eliminated all primary deuterons, but was penetrable for fast fusion protons. The foil and first PM-355 detector were penetrable for fast neutrons, which were converted into recoil-protons in the polyethylene and recorded in the second PM-355 detector. The “sandwiches” were irradiated by discharges of comparable neutron-yields. Analyses of etched tracks and computer simulations of the fusion-products behavior in the detectors were performed.

  14. Response of ISSEC protected first walls to DT and DD plasma neutrons

    International Nuclear Information System (INIS)

    Avci, H.I.; Kulcinski, G.L.

    1976-01-01

    It has been demonstrated that the displacement damage and gas production rates can be reduced in CTR first walls by employing passive carbon shields. Reductions in displacement damage range from 3 to 5 for 12.5 cm shield thickness and from 7 to 14 in gas production rates with the same carbon thickness. The factors of reduction are 8 to 20 for the displacements and 17 to 80 for the gas production if a 25 cm shield is used. Depending on whether the isotopes causing the radioactivity are produced as a result of fast or thermal neutron activation, the first wall radioactivity can either go up or down with the increasing carbon shield thickness. It has been found that at shutdown radioactivity in 316 SS, Al, and Nb first walls is reduced with increasing carbon thickness while the activities in V and Ta are increased. Long term radioactivity displays the same trends in Al, 316 SS and Ta as short term radioactivity. However, the long term activity in Nb increases and that in V decreases with increasing shield thickness. It has also been found that systems operating on a D-D plasma cycle have higher displacement rates than respective D-T cycle systems. Gas production rates are slightly lower in D-D systems except for He production in 316 SS. This is due to the higher 59 Ni (n,α) cross sections for thermal neutrons

  15. From microscopic to macroscopic dynamics in mean-field theory: effect of neutron skin on fusion barrier and dissipation

    Energy Technology Data Exchange (ETDEWEB)

    Lacroix, D

    2001-07-01

    In this work, we introduce a method to reduce the microscopic mean-field theory to a classical macroscopic dynamics at the initial stage of fusion reaction. We show that TDHF (Time-dependent Hartree-Fock) could be a useful tool to infer information on the fusion barrier as well as on one-body dissipation effect. We apply the reduction of information to the case of head-on reaction between a {sup 16}O and {sup 16,22,24,28}O in order to quantify the effect of neutron skin on fusion. We show that the precise determination of fusion barrier requires, in addition to the relative distance between center of mass, the introduction of an additional collective coordinate that explicitly breaks the neutron-proton symmetry. With this additional collective variable, we obtain a rather precise determination of the barrier position, height and diffuseness as well as one-body friction. (author)

  16. Liquid Wall Chambers

    Energy Technology Data Exchange (ETDEWEB)

    Meier, W R

    2011-02-24

    The key feature of liquid wall chambers is the use of a renewable liquid layer to protect chamber structures from target emissions. Two primary options have been proposed and studied: wetted wall chambers and thick liquid wall (TLW) chambers. With wetted wall designs, a thin layer of liquid shields the structural first wall from short ranged target emissions (x-rays, ions and debris) but not neutrons. Various schemes have been proposed to establish and renew the liquid layer between shots including flow-guiding porous fabrics (e.g., Osiris, HIBALL), porous rigid structures (Prometheus) and thin film flows (KOYO). The thin liquid layer can be the tritium breeding material (e.g., flibe, PbLi, or Li) or another liquid metal such as Pb. TLWs use liquid jets injected by stationary or oscillating nozzles to form a neutronically thick layer (typically with an effective thickness of {approx}50 cm) of liquid between the target and first structural wall. In addition to absorbing short ranged emissions, the thick liquid layer degrades the neutron flux and energy reaching the first wall, typically by {approx}10 x x, so that steel walls can survive for the life of the plant ({approx}30-60 yrs). The thick liquid serves as the primary coolant and tritium breeding material (most recent designs use flibe, but the earliest concepts used Li). In essence, the TLW places the fusion blanket inside the first wall instead of behind the first wall.

  17. Fiscal year 1976T (add-on quarter) DT fusion neutron irradiations and dosimetry at the LLL rotating target neutron source

    International Nuclear Information System (INIS)

    MacLean, S.C.

    1977-01-01

    The DT fusion neutron irradiation of more than 90 samples during seven irradiation periods (beam-on time of more than 430.9 hours) is described. Experiments from 15 individuals representing six institutions are summarized. The numbers of UCID dosimetry reports detailing each of the irradiations is given

  18. An overview of the development of the first wall and other principal components of a laser fusion power plant

    International Nuclear Information System (INIS)

    Sethian, John D.; Raffray, A. Rene; Latkowski, Jeffery; Blanchard, James P.; Snead, Lance; Renk, Timothy J.; Sharafat, Shahram

    2005-01-01

    This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations

  19. An overview of the development of the first wall and other principal components of a laser fusion power plant

    Science.gov (United States)

    Sethian, John D.; Raffray, A. Rene; Latkowski, Jeffery; Blanchard, James P.; Snead, Lance; Renk, Timothy J.; Sharafat, Shahram

    2005-12-01

    This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.

  20. An overview of the development of the first wall and other principal components of a laser fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Sethian, John D. [Plasma Physics Division, Naval Research Laboratory, 4555 Overlook Av. SW, Washington, DC 20375 (United States)]. E-mail: sethian@this.nrl.navy.mil; Raffray, A. Rene [University of California, San Diego, La Jolla, CA 92093 (United States); Latkowski, Jeffery [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Blanchard, James P. [University of Wisconsin, Madison, WI 53706 (United States); Snead, Lance [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Renk, Timothy J. [Sandia National Laboratory, Albuquerque, NM 87185 (United States); Sharafat, Shahram [University of California, Los Angeles, Los Angeles, CA 90095 (United States)

    2005-12-15

    This paper introduces the JNM Special Issue on the development of a first wall for the reaction chamber in a laser fusion power plant. In this approach to fusion energy a spherical target is injected into a large chamber and heated to fusion burn by an array of lasers. The target emissions are absorbed by the wall and encapsulating blanket, and the resulting heat converted into electricity. The bulk of the energy deposited in the first wall is in the form of X-rays (1.0-100 keV) and ions (0.1-4 MeV). In order to have a practical power plant, the first wall must be resistant to these emissions and suffer virtually no erosion on each shot. A wall candidate based on tungsten armor bonded to a low activation ferritic steel substrate has been chosen as the initial system to be studied. The choice was based on the vast experience with these materials in a nuclear environment and the ability to address most of the key remaining issues with existing facilities. This overview paper is divided into three parts. The first part summarizes the current state of the development of laser fusion energy. The second part introduces the tungsten armored ferritic steel concept, the three critical development issues (thermo-mechanical fatigue, helium retention, and bonding) and the research to address them. Based on progress to date the latter two appear to be resolvable, but the former remains a challenge. Complete details are presented in the companion papers in this JNM Special Issue. The third part discusses other factors that must be considered in the design of the first wall, including compatibility with blanket concepts, radiological concerns, and structural considerations.

  1. Acceleration techniques for the direct use of CAD-based geometry in fusion neutronics analysis

    International Nuclear Information System (INIS)

    Wilson, Paul P.H.; Tautges, Timothy J.; Kraftcheck, Jason A.; Smith, Brandon M.; Henderson, Douglass L.

    2010-01-01

    The Direct Accelerated Geometry Monte Carlo (DAGMC) software library offers a unique approach to performing neutronics analysis on CAD-based geometries of fusion systems. By employing a number of acceleration techniques, the ray-tracing operations that are fundamental to Monte Carlo radiation transport are implemented efficiently for direct use on the CAD-based solid model, eliminating the need to translate to the native Monte Carlo input language. By forming hierarchical trees of oriented bounding boxes, one for each facet that results from a high-fidelity tessellation of the model, the ray-tracing performance is adequate to permit detailed analysis of large complex systems. In addition to the reduction in human effort and improvement in quality assurance that is found in the translation approaches, the DAGMC approach also permits the analysis of geometries with higher order surfaces that cannot be represented by many native Monte Carlo radiation transport tools. The paper describes the various acceleration techniques and demonstrates the resulting capability in a real fusion neutronics analysis.

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  3. Preliminary design of GDT-based 14 MeV neutron source

    International Nuclear Information System (INIS)

    Du Hongfei; Chen Dehong; Wang Hui; Wang Fuqiong; Jiang Jieqiong; Wu Yican; Chen Yiping

    2012-01-01

    To meet the need of D-T fusion neutron source for fusion material testing, design goals were presented in this paper according to the international requirements of neutron source for fusion material testing. A preliminary design scheme of GDT-based 14 MeV neutron source was proposed, and a physics model of the neutron source was built based on progress of GDT experiments. Two preliminary design schemes (i. e. FDS-GDT1, FDS-GDT2) were designed; among which FDS-GDT2 can be used for fusion material testing with neutron first wall loading of 2 MW/m 2 . (authors)

  4. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

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

  5. A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

    International Nuclear Information System (INIS)

    Morgan, L.W.G.; Packer, L.W.

    2014-01-01

    Highlights: • The issue of a potentially producing a large tritium surplus inventory, within a solid breeder, is addressed. • A possible solution to this problem is presented in the form of a neutron poison based tritium production controller. • The tritium surplus inventory has been modelled by the FATI code for a simplified WCCB model and as a function of time. • It has been demonstrated that the tritium surplus inventory can be managed, which may impact on safety considerations. - Abstract: The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist. This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into

  6. A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, L.W.G., E-mail: Lee.Morgan@CCFE.ac.uk; Packer, L.W.

    2014-10-15

    Highlights: • The issue of a potentially producing a large tritium surplus inventory, within a solid breeder, is addressed. • A possible solution to this problem is presented in the form of a neutron poison based tritium production controller. • The tritium surplus inventory has been modelled by the FATI code for a simplified WCCB model and as a function of time. • It has been demonstrated that the tritium surplus inventory can be managed, which may impact on safety considerations. - Abstract: The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist. This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into

  7. Nuclear-heat deposition for a fusion-like neutron environment

    International Nuclear Information System (INIS)

    Carter, L.L.; Hegberg, D.E.; Wilcox, A.D.

    1981-10-01

    Calculated nuclear heat deposition profiles within the thermal shield of the FMIT facility are sensitive to the cross-section data base - particularly an energy balance consistency between gamma production cross-sections and neutron KERMA factors. Infinite medium calculations were made with the Monte Carlo code to provide integral validations of energy balances relevant to this aspect of the data base. Inconsistencies were found and corrected. There was also concern about the adequacy of the high energy cross sections (10 MeV < E < 30 MeV) for the moderation and transport of the (d,Li) source neutrons. A preliminary analysis of a measurement with a (d,Li) source in the center of an iron block has improved our confidence in the high energy cross section - data base for this application. Monte Carlo calculations have been utilized to calculate three-dimensional profiles of nuclear heat deposition. Representative profiles were displayed for two walls of the FMIT test cell

  8. ZZ FENDL-2, Evaluated Nuclear Data Library for Fusion Neutronics Applications

    International Nuclear Information System (INIS)

    2005-01-01

    Description: FENDL: Fusion Evaluated Nuclear Data Library. Materials/nuclides: H 1 , H 2 , H 3 , He 3 , He 4 , Li 6 , Li 7 , Be 9 , B 10 , B 11 , C 12 , N 14 , N 15 , O 16 , F 19 , Na 23 , Mg nat , Al 27 , Si 28 , Si 29 , Si 30 , P 31 , S nat , Cl 35 , Cl 37 , K nat , Ca nat , Ti 46 , Ti 47 , Ti 48 , Ti 49 , Ti 50 , V nat , Cr 50 , Cr 52 , Cr 53 , Cr 54 , Mn 55 , Fe 54 , Fe 56 , Fe 57 , Fe 58 , Co 59 , Ni 58 , Ni 60 , Ni 61 , Ni 62 , Ni 64 , Cu 63 , Cu 65 , Ga nat , Zr nat , Nb 93 , Mo 92 , Mo 94 , Mo 95 , Mo 96 , Mo 97 , Mo 98 , Mo 100 , Sn nat , Ta 181 , W 182 , W 183 , W 184 , W 186 , Au 197 , Pb 206 , Pb 207 , Pb 208 , Bi 209 . Photo-atomic data. IAEA1364/02: FENDL version 2.0 consists of the following sub-libraries: - ACTIVATION (FENDL/A-2.0)- neutron activation cross sections for 13006 reactions on 739 targets ranging from 1-H up to 248-Cm at incident energies up to 20 MeV. Pointwise and processed data in different formats are included. Plots are available. - DECAY (FENDL/D-2.0) - decay properties (decay type, decay energy, and half life) for 1867 nuclides and isomers. FENDL/D-2.0 sub-library is complementary to the activation sub-library. Pointwise and processed data are included. - DOSIMETRY (FENDL/DS-2.0) - neutron cross sections to be used for reactor neutron dosimetry by foil activation, radiation damage cross-sections, and benchmark neutron spectra. This sub-library is identical to the International Reactor Dosimetry File (IRDF-90). Pointwise and processed data are included. - FUSION (FENDL/C-2.0) - charged-particle cross sections for the following fusion reactions: 1-H 2 (d,n)2-He 3 , 1-H 2 (d,p)1-H 3 , 2-He 3 (d,p)2-He 4 , 1-H-3(t,2n)2-He 4 , and 1-H 3 (d,n)2-He 4 . Pointwise and processed data are included. - TRANSPORT - validated basic nuclear data (neutron-nucleus interaction including photon production, and photon-atom interaction cross sections) for 57 nuclides relevant for fusion. In addition to the pointwise data (FENDL/E-2.0), the sub

  9. Materials for fusion reactors

    International Nuclear Information System (INIS)

    Ehrlich, K.; Kaletta, D.

    1978-03-01

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

  10. Neutron spectroscopy measurements of 14 MeV neutrons at unprecedented energy resolution and implications for deuterium-tritium fusion plasma diagnostics

    Science.gov (United States)

    Rigamonti, D.; Giacomelli, L.; Gorini, G.; Nocente, M.; Rebai, M.; Tardocchi, M.; Angelone, M.; Batistoni, P.; Cufar, A.; Ghani, Z.; Jednorog, S.; Klix, A.; Laszynska, E.; Loreti, S.; Pillon, M.; Popovichev, S.; Roberts, N.; Thomas, D.; Contributors, JET

    2018-04-01

    An accurate calibration of the JET neutron diagnostics with a 14 MeV neutron generator was performed in the first half of 2017 in order to provide a reliable measurement of the fusion power during the next JET deuterium-tritium (DT) campaign. In order to meet the target accuracy, the chosen neutron generator has been fully characterized at the Neutron Metrology Laboratory of the National Physical Laboratory (NPL), Teddington, United Kingdom. The present paper describes the measurements of the neutron energy spectra obtained using a high-resolution single-crystal diamond detector (SCD). The measurements, together with a new neutron source routine ‘ad hoc’ developed for the MCNP code, allowed the complex features of the neutron energy spectra resulting from the mixed D/T beam ions interacting with the T/D target nuclei to be resolved for the first time. From the spectral analysis a quantitative estimation of the beam ion composition has been made. The unprecedented intrinsic energy resolution (<1% full width at half maximum (FWHM) at 14 MeV) of diamond detectors opens up new prospects for diagnosing DT plasmas, such as, for instance, the possibility to study non-classical slowing down of the beam ions by neutron spectroscopy on ITER.

  11. Powder metallurgical processing of self-passivating tungsten alloys for fusion first wall application

    International Nuclear Information System (INIS)

    López-Ruiz, P.; Ordás, N.; Iturriza, I.; Walter, M.; Gaganidze, E.; Lindig, S.; Koch, F.; García-Rosales, C.

    2013-01-01

    Self-passivating tungsten based alloys are expected to provide a major safety advantage compared to pure tungsten, presently the main candidate material for first wall armour of future fusion reactors. In case of a loss of coolant accident with simultaneous air ingress, a protective oxide scale will be formed on the surface of W avoiding the formation of volatile and radioactive WO 3 . Bulk WCr12Ti2.5 alloys were manufactured by mechanical alloying (MA) and hot isostatic pressing (HIP), and their properties compared to bulk WCr10Si10 alloys from previous work. The MA parameters were adjusted to obtain the best balance between lowest possible amount of contaminants and effective alloying of the elemental powders. After HIP, a density >99% is achieved for the WCr12Ti2.5 alloy and a very fine and homogeneous microstructure with grains in the submicron range is obtained. Unlike the WCr10Si10 material, no intergranular ODS phase inhibiting grain growth was detected. The thermal and mechanical properties of the WCr10Si10 material are dominated by the silicide (W,Cr) 5 Si 3 ; it shows a sharp ductile-to brittle transition in the range 1273–1323 K. The thermal conductivity of the WCr12Ti2.5 alloy is close to 50 W/mK in the temperature range of operation; it exhibits significantly higher strength and lower DBTT – around 1170 K – than the WCr10Si10 material

  12. Investigation of cascade-type falling liquid-film along first wall of laser-fusion reactor

    International Nuclear Information System (INIS)

    Kunugi, T.; Nakai, T.; Kawara, Z.; Norimatsu, T.; Kozaki, Y.

    2008-01-01

    To protect the first wall of an inertia fusion reactor from extremely high heat flux, X-rays, alpha particles and fuel debris caused by a nuclear fusion reaction, a 'cascade-type' falling liquid-film flow is proposed as a 'liquid-wall' concept. The flow visualization experiment to investigate the feasibility of this liquid-wall concept has been conducted. The preliminary numerical simulation results suggest that the current cascade structure design should be improved because less thermal-mixing is expected. The cascade-type structure has, therefore, been redesigned. This new cascade-type first wall consists of a liquid reservoir which has a free-surface to maintain a constant water head in the rear, and connects to a slit composed of two plates, i.e., the first wall is connected to a slit which is partially made up of the first wall to begin with it. The numerical simulations were performed on the new cascade-type first wall and they show the stable liquid-film flow on it. Moreover, the POP (proof-of-principle) flow visualization experiments, which satisfy the Weber number coincident condition, are carried out using water as the working fluid. By comparing the numerical and experimental results, it was found that the liquid-film flow with 3-5 mm thickness could be stably established. According to these results for the new cascade-type first wall concept, it was confirmed that the coolant flow rate and the thickness of the liquid-film could be controlled if the Weber number coincident condition was satisfied

  13. Positron annihilation lifetime measurements of vanadium alloy and F82H irradiated with fission and fusion neutrons

    International Nuclear Information System (INIS)

    Sato, K.; Inoue, K.; Yoshiie, T.; Xu, Q.; Wakai, E.; Kutsukake, C.; Ochiai, K.

    2009-01-01

    V-4Cr-4Ti, F82H, Ni and Cu were irradiated with fission and fusion neutrons at room temperature and 473 K. Defect structures were analyzed and compared using positron annihilation lifetime measurement, and microstructural evolution was discussed. The mean lifetime of positrons (the total amount of residual defects) increased with the irradiation dose. The effect of cascade impact was detected in Ni at room temperature. The size and the number of vacancy clusters were not affected by the displacement rate in the fission neutron irradiation at 473 K for the metals studied. The vacancy clusters were not formed in V-4Cr-4Ti irradiated at 473 K in the range of 10 -6 -10 -3 dpa. In F82H irradiated at 473 K, the defect evolution was prevented by pre-existing defects. The mean lifetime of positrons in fission neutron irradiation was longer than that in fusion neutron irradiation in V-4Cr-4Ti at 473 K. It was interpreted that more closely situated subcascades were formed in the fusion neutron irradiation and subcascades interacted with each other, and consequently the vacancy clusters did not grow larger.

  14. Progress on using deuteron-deuteron fusion generated neutrons for 40Ar/39Ar sample irradiation

    Science.gov (United States)

    Rutte, Daniel; Renne, Paul R.; Becker, Tim; Waltz, Cory; Ayllon Unzueta, Mauricio; Zimmerman, Susan; Hidy, Alan; Finkel, Robert; Bauer, Joseph D.; Bernstein, Lee; van Bibber, Karl

    2017-04-01

    We present progress on the development and proof of concept of a deuteron-deuteron fusion based neutron generator for 40Ar/39Ar sample irradiation. Irradiation with deuteron-deuteron fusion neutrons is anticipated to reduce Ar recoil and Ar production from interfering reactions. This will allow dating of smaller grains and increase accuracy and precision of the method. The instrument currently achieves neutron fluxes of ˜9×107 cm-2s-1 as determined by irradiation of indium foils and use of the activation reaction 115In(n,n')115mIn. Multiple foils and simulations were used to determine flux gradients in the sample chamber. A first experiment quantifying the loss of 39Ar is underway and will likely be available at the time of the presentation of this abstract. In ancillary experiments via irradiation of K salts and subsequent mass spectrometric analysis we determined the cross-sections of the 39K(n,p)39Ar reaction at ˜2.8 MeV to be 160 ± 35 mb (1σ). This result is in good agreement with bracketing cross-section data of ˜96 mb at ˜2.45 MeV and ˜270 mb at ˜4 MeV [Johnson et al., 1967; Dixon and Aitken, 1961 and Bass et al. 1964]. Our data disfavor a much lower value of ˜45 mb at 2.59 MeV [Lindström & Neuer, 1958]. In another ancillary experiment the cross section for 39K(n,α)36Cl at ˜2.8 MeV was determined as 11.7 ± 0.5 mb (1σ), which is significant for 40Ar/39Ar geochronology due to subsequent decay to 36Ar as well as for the determination of production rates of cosmogenic 36Cl. Additional experiments resolving the cross section functions on 39K between 1.5 and 3.6 MeV are on their way using the LICORNE neutron source of the IPN Orsay tandem accelerator. Results will likely be available at the time of the presentation of this abstract. While the neutron generator is designed for fluxes of ˜109 cm-2s-1, arcing in the sample chamber currently limits the power—straightforwardly correlated to the neutron flux—the generator can safely be run at. Further

  15. Dynamic response of thermal neutron measurements in electrochemically produced cold fusion subject to pulsed current

    International Nuclear Information System (INIS)

    Granada, Jose; Converti, Jose; Mayer, Roberto; Guido, German; Florido, Pablo; Patino, Nestor; Sobehart, Leonardo; Gomez, Silvia; Larreteguy, Axel

    1988-01-01

    The present work shows the results of measurements performed on electrolytic cells using a high efficiency (22%) neutron detection system in combination with a procedure involving a non-stationary current through the cell's circuit. Cold fusion was produced in electrolytic cells containing LiH dissolved in heavy water with a palladium cathode. The dynamic response to low frequency current pulses was measured. Characteristic patterns showing one or two bumps were obtained in a repeatable fashion. These patterns are strongly dependent on the previous charging history of the cathode. The technique employed seems to be very convenient as a research tool for a systematic study of the different variables governing the phenomenon. (Author)

  16. Analysis of fusion neutronics calculations and appraisal of UW cross-section library

    International Nuclear Information System (INIS)

    Xie Jianping; Li Xingzhong; Ying Chuntong

    1989-01-01

    A series of calculations for different cases (especially for the values of tritium breeding ratio T, and the fuel breeding ratio F in the blanket of a hybrid reactor) were carried out by using ANISN program and UW cross-section library. The comparison with other results in China and abroad kalso was done. It was shownwn that the installation and execution of ANISN program on ELXSI machine at Tsinghua University are successful, and the UW cross-section library is reliable. It may be used for fusion neutronics calculation in the future. The paper also points out that the difference between the calculations and by the authors are due to jthe different in cross-section data used

  17. Fast rigorous numerical method for the solution of the anisotropic neutron transport problem and the NITRAN system for fusion neutronics application. Pt. 2

    International Nuclear Information System (INIS)

    Takahashi, A.; Rusch, D.

    1979-10-01

    The I*-method, which is a non-approximative treatment of the neutron balance equations by the use of double-differential cross sections and a generalized angular transfer probability, is realized within the NITRAN system. It is shown, by means of test calculations for assemblies related to fusion reactor neutronics that double-differential cross section data provide substantial progress in transport problems with kinematically complicated reaction channels like (n,2n), (n,n'γ), and (n,n'α), because the I*-method is free from kinematic assumptions. The properties of the exponential method to generate the supplementary equations to the SN equations are investigated. (orig.) [de

  18. Measurement and analysis of 14 MeV neutron-induced double-differential neutron emission cross sections needed for fission and fusion reactor technology

    International Nuclear Information System (INIS)

    Wang Dahai; Mehta, M.K.

    1988-07-01

    The main objectives of this IAEA Co-ordinated Research Programme are to improve the current status of data for 14 MeV neutron-induced double-differential neutron emission cross sections for V, Cr, Fe, Nb, Ta and 238 U. The principal objectives of this first meeting were to report on the status of participants' work, to exchange experience in experimental work and to establish the future work. Considering the unsatisfactory status of the data for 6 Li, 7 Li, 9 Be, Mo, W and Bi and their importance in fusion reactor technology participants agreed to include these isotopes in the programme

  19. Integral test of KERMA data for SS304 stainless steel in the D-T fusion neutron environment

    International Nuclear Information System (INIS)

    Ikeda, Y.; Kosako, K.; Konno, C.

    1994-01-01

    The KERMA (Kinetic Energy Release Material) data play the fundamental role for estimating nuclear heating in the structural components of fusion reactors. The data are produced from the large body of nuclear data relevant to reaction channels associated with the kinetic energy release. Both contributions by neutron and gamma-ray should be addressed to arrived at the final heating products. Extensive efforts have been devoted to the neutron and γ-ray transport profile in many materials, resulting in the validation of cross section data. However, the experimental verification of KERMA data, which is a highly integrated product of neutron and γ-ray, has been limited from the lack of available experimental data. Through the JAERI/USDOE collaborative program on fusion neutronics, novel experimental technique for the direct nuclear heating due to 14 MeV neutrons has been developed based on a micro calorimetric system. The technique demonstrated excellent capability for detecting the temperature rise due to nuclear heating and pertinent verification for the calculation data and methods. This paper deals with the most recent experimental endeavor for the direct nuclear heating measurement in SS-304 stainless steel assembly, where appreciably large amounts of slow neutron and associated secondary γ-rays dominated the field. The nuclear heating up to 200 mm depth in the SS-304 assembly were derived from detected temperature rise employing large SS-304 block type probe materials

  20. Neutronics analysis of water-cooled energy production blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Jiang Jieqiong; Wang Minghuang; Chen Zhong; Qiu Yuefeng; Liu Jinchao; Bai Yunqing; Chen Hongli; Hu Yanglin

    2010-01-01

    Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor (M) and tritium breeding ratio (TBR) in a fusion-fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

  1. Advanced fusion reactor

    International Nuclear Information System (INIS)

    Tomita, Yukihiro

    2003-01-01

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

  2. Advanced fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-04-01

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

  3. Comprehensive neutron cross-section and secondary energy distribution uncertainty analysis for a fusion reactor

    International Nuclear Information System (INIS)

    Gerstl, S.A.W.; LaBauve, R.J.; Young, P.G.

    1980-05-01

    On the example of General Atomic's well-documented Power Generating Fusion Reactor (PGFR) design, this report exercises a comprehensive neutron cross-section and secondary energy distribution (SED) uncertainty analysis. The LASL sensitivity and uncertainty analysis code SENSIT is used to calculate reaction cross-section sensitivity profiles and integral SED sensitivity coefficients. These are then folded with covariance matrices and integral SED uncertainties to obtain the resulting uncertainties of three calculated neutronics design parameters: two critical radiation damage rates and a nuclear heating rate. The report documents the first sensitivity-based data uncertainty analysis, which incorporates a quantitative treatment of the effects of SED uncertainties. The results demonstrate quantitatively that the ENDF/B-V cross-section data files for C, H, and O, including their SED data, are fully adequate for this design application, while the data for Fe and Ni are at best marginally adequate because they give rise to response uncertainties up to 25%. Much higher response uncertainties are caused by cross-section and SED data uncertainties in Cu (26 to 45%), tungsten (24 to 54%), and Cr (up to 98%). Specific recommendations are given for re-evaluations of certain reaction cross-sections, secondary energy distributions, and uncertainty estimates

  4. Fusion

    CERN Document Server

    Mahaffey, James A

    2012-01-01

    As energy problems of the world grow, work toward fusion power continues at a greater pace than ever before. The topic of fusion is one that is often met with the most recognition and interest in the nuclear power arena. Written in clear and jargon-free prose, Fusion explores the big bang of creation to the blackout death of worn-out stars. A brief history of fusion research, beginning with the first tentative theories in the early 20th century, is also discussed, as well as the race for fusion power. This brand-new, full-color resource examines the various programs currently being funded or p

  5. Radiography studies with gamma rays produced by 14-MeV fusion neutrons

    International Nuclear Information System (INIS)

    Smith, D.L.; Ikeda, Yujiro; Uno, Yoshitomo

    1996-01-01

    Oxygen contained in pure water has been activated via the 16 O(n, p) 16 N reaction using 14-MeV neutrons produced at a neutron generator with the 3 H(d,n) 4 He source. Photons of 6.129 and 7.115 MeV, generated by the decay of 7.13-second 16 N, were then used to demonstrate the feasibility of employing highly penetrating, nearly monoenergetic gamma rays for radiography studies of thick, dense objects composed of elements with medium to relatively high atomic numbers. A simple radiography apparatus was constructed by circulating water continuously between a position near the target of the neutron generator and a remote location where photon transmission measurements were conducted. A sodium iodide scintillator was employed to detect the photons. Pulses equivalent to photon energies smaller than 2.506 MeV (corresponding to the cascade sum of 1.333- and 1.173-MeV gamma rays from the decay of 5.271-year 60 Co) were rejected by the electronics settings in order to reduce background and improve the signal-to-noise (S/N) ratio. Respectable S/N ratios on the order of 20-to-1 were achieved with this setup. Most of the background (N) could be attributed to ambient environmental radiation and cosmic-ray interactions with the lead shielding and detector. Four representative objects were examined by photon radiography in this study. This demonstrated how such - interesting features as hidden holes and discontinuities in atomic number could be easily identified from observed variations in the intensity of transmitted photons. Some advantages of this technique are described, and potential applications are suggested for a future scenario where fusion reactors are used to generate electric power and very intense sources of high-energy photons from 16 N decay are continuously available as a byproduct of the reactor cooling process

  6. Survey on neutron pre-emission at the fusion of 11 Li halo nuclei with light targets

    International Nuclear Information System (INIS)

    Petrascu, M.

    1999-01-01

    The neutron halo nuclei characterized by very large matter radii, small separation energy and small internal momentum of the valence neutrons, were discovered by Tanihata and co-workers. Until now, the halo nuclei were investigated mostly by elastic, inelastic scattering and breakup processes. It was recently predicted, that due to the very large dimension of 11 Li, one may expect, that in a fusion experiment on a light target, the valence neutrons will not be absorbed together with the 9 Li core, but will be emitted in the early stage of the reaction process. The first experiment aiming to check this expectation, was performed at the RIKEN-RIPS facility. In the experimental setup, the MUSIC chamber, achieved in the frame of IFIN-HH - RIKEN cooperation, played an important role. The obtained results confirm the prediction, indicating to a novel pre-emission effect (near 40 %, for one or two neutrons). The pre-emission of neutron pairs was investigated by time-position coincidences. It is considered that the pre-emission of neutron pairs is responsible for the experimentally observed strong neutron focusing effect. An experiment aiming to a large number of n-n coincidences, based on a new neutron array detector built in the frame of IFIN-HH - RIKEN cooperation, is in preparation. A model for the pre-emission probabilities calculations has been also worked out. Good agreement with the experimental data has been obtained. (author)

  7. The longitudinal wall impedance instability in a heavy-ion fusion driver

    International Nuclear Information System (INIS)

    Callahan, D.A.; Langdon, A.B.; Friedman, A.; Haber, I.

    1997-01-01

    For more than ten years [J. Bisognano, I. Haber, L. Smith, IEEE Trans. Nucl. Sci. NS-30, 2501 (1983)], the longitudinal wall impedance instability was thought to be a serious threat to the success of heavy-ion driven inertial confinement fusion. This instability is a open-quotes resistive wallclose quotes instability, driven by the impedance of the induction modules used to accelerate the beam. Early estimates of the instability growth rate predicted tens of e-folds due to the instability which would modulate the current and increase the longitudinal momentum spread and prevent focusing the ion beam on the small spot needed at the target. We have simulated this instability using an r-z particle-in-cell code which includes a model for the module impedance. These simulations, using driver parameters, show that growth due to the instability is smaller than in previous calculations. We have seen that growth is mainly limited to one head to tail transit by a space-charge wave. In addition, the capacitive component of the module impedance, which was neglected in the early work of Lee [E. P. Lee, Proc. Linear Accelerator Conference, (UCRL-86452), Santa Fe, NM, 1981] significantly reduces the growth rate. We have also included in the simulation intermittently applied axial confining fields which are thought to be the major source of perturbations to seed the longitudinal instability. Simulations show the beam can adjust to a systematic error in the longitudinal confining fields while a random error excites the most unstable wavelength of the instability. These simulations show that the longitudinal instability must be taken into account in a driver design, but it is not the major factor it was once thought to be. copyright 1997 American Institute of Physics

  8. Powder metallurgical processing of self-passivating tungsten alloys for fusion first wall application

    Energy Technology Data Exchange (ETDEWEB)

    López-Ruiz, P.; Ordás, N.; Iturriza, I. [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain); Walter, M.; Gaganidze, E. [Karlsruhe Institute of Technology (KIT), D-76344 Eggenstein-Leopoldshafen (Germany); Lindig, S.; Koch, F. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); García-Rosales, C., E-mail: cgrosales@ceit.es [CEIT and Tecnun (University of Navarra), E-20018 San Sebastian (Spain)

    2013-11-15

    Self-passivating tungsten based alloys are expected to provide a major safety advantage compared to pure tungsten, presently the main candidate material for first wall armour of future fusion reactors. In case of a loss of coolant accident with simultaneous air ingress, a protective oxide scale will be formed on the surface of W avoiding the formation of volatile and radioactive WO{sub 3}. Bulk WCr12Ti2.5 alloys were manufactured by mechanical alloying (MA) and hot isostatic pressing (HIP), and their properties compared to bulk WCr10Si10 alloys from previous work. The MA parameters were adjusted to obtain the best balance between lowest possible amount of contaminants and effective alloying of the elemental powders. After HIP, a density >99% is achieved for the WCr12Ti2.5 alloy and a very fine and homogeneous microstructure with grains in the submicron range is obtained. Unlike the WCr10Si10 material, no intergranular ODS phase inhibiting grain growth was detected. The thermal and mechanical properties of the WCr10Si10 material are dominated by the silicide (W,Cr){sub 5}Si{sub 3}; it shows a sharp ductile-to brittle transition in the range 1273–1323 K. The thermal conductivity of the WCr12Ti2.5 alloy is close to 50 W/mK in the temperature range of operation; it exhibits significantly higher strength and lower DBTT – around 1170 K – than the WCr10Si10 material.

  9. Neutron metrology in LAMPF, USA

    International Nuclear Information System (INIS)

    Ketema, D.J.; Nolthenius, H.J.

    1990-08-01

    The characterization of appropriate materials for fusion reactors requires a high intensity neutron source which simulates the neutron spectrum and the radiation conditions at the positions of interest in a fusion reactor (first wall). A neutron spectrum of interest is found in the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF). Various ceramic materials and some polycrystalline graphites were irradiated in this facility during two intervals of time in 1986 and 1987. The specimens were accompanied by several sets with activation detectors. This report presents the saturation activities per atom obtained from these sets. (author). 3 refs.; 8 figs.; 10 tabs

  10. Report of the study meeting on the interaction between plasma and the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Miyahara, Akira; Akaishi, Kenya; Kawamura, Takaichi; Kabetani, Zenzaburo; Sagara, Akio.

    1978-12-01

    The study meeting on the interaction between plasma and the first wall of a fusion reactor was held from July 24 to July 27, 1978. At this meeting, discussions were made on the interaction between plasma and wall and the effect of impurities. Reports on the ISS observation concerning the Mo surface as a limiter, on the measurement of sputter rate by a microbalance, on the surface roughness of the materials for the first wall at the atomic order, on the selective sputtering of binary alloys, and on the physical and chemical sputtering on the material surface of C and SiC were also presented. The research projects of the Institute of Plasma Physics and Hokkaido University were introduced. Collaboration of two groups was considered. (Kato, T.)

  11. Smart tungsten alloys as a material for the first wall of a future fusion power plant

    Science.gov (United States)

    Litnovsky, A.; Wegener, T.; Klein, F.; Linsmeier, Ch.; Rasinski, M.; Kreter, A.; Unterberg, B.; Coenen, J. W.; Du, H.; Mayer, J.; Garcia-Rosales, C.; Calvo, A.; Ordas, N.

    2017-06-01

    Tungsten is currently deemed as a promising plasma-facing material (PFM) for the future power plant DEMO. In the case of an accident, air can get into contact with PFMs during the air ingress. The temperature of PFMs can rise up to 1200 °C due to nuclear decay heat in the case of damaged coolant supply. Heated neutron-activated tungsten forms a volatile radioactive oxide which can be mobilized into the atmosphere. New self-passivating ‘smart’ alloys can adjust their properties to the environment. During plasma operation the preferential sputtering of lighter alloying elements will leave an almost pure tungsten surface facing the plasma. During an accident the alloying elements in the bulk are forming oxides thus protecting tungsten from mobilization. Good plasma performance and the suppression of oxidation are required for smart alloys. Bulk tungsten (W)-chroimum (Cr)-titanium (Ti) alloys were exposed together with pure tungsten (W) samples to the steady-state deuterium plasma under identical conditions in the linear plasma device PSI 2. The temperature of the samples was ~576 °C-715 °C, the energy of impinging ions was 210 eV matching well the conditions expected at the first wall of DEMO. Weight loss measurements demonstrated similar mass decrease of smart alloys and pure tungsten samples. The oxidation of exposed samples has proven no effect of plasma exposure on the oxidation resistance. The W-Cr-Ti alloy demonstrated advantageous 3-fold lower mass gain due to oxidation than that of pure tungsten. New yttrium (Y)-containing thin film systems are demonstrating superior performance in comparison to that of W-Cr-Ti systems and of pure W. The oxidation rate constant of W-Cr-Y thin film is 105 times less than that of pure tungsten. However, the detected reactivity of the bulk smart alloy in humid atmosphere is calling for a further improvement.

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

    International Nuclear Information System (INIS)

    Dierckx, R.; Emigh, C.R.

    1977-01-01

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

  13. Neutron activation analysis of heavy metal binding by fungal cell walls

    International Nuclear Information System (INIS)

    Crusberg, T.C.; Mayer, J.A.

    1994-01-01

    Aqueous effluents are produced during nuclear power and nuclear weapons development activities which frequently contain low levels of dissolved radioactive nuclides. A number of laboratories are now focusing attention to renewable biological materials to provide traps for low concentrations of dissolved radioactive metal ions in wastewater effluents. The term BIOTRAP can be used to describe such materials, and in this laboratory cell wall preparations of the fungus Penicillium ochro-chloron have been employed to demonstrate their capacity and affinity to reversibly bind and remove copper(2). Since neutron activation analysis (NAA) was readily available, that method was one of several applied to this problem as a suitable analytical methodology to study heavy metal-to-BIOTRAP interactions. Copper and mercury provide good examples of metals which are capable of undergoing activation by thermal neutrons. In NAA, 63 Cu (69.1% natural abundance) is converted to 64 Cu which has a half live of 12.7 hr, and 202 Hg (29.7 % natural abundance) is converted to 203 Hg which has a half life of 46.,6 d

  14. Photon and neutron dose discrimination using low pressure proportional counters with graphite and A150 walls

    International Nuclear Information System (INIS)

    Kylloenen, J.; Lindborg, L.

    2005-01-01

    Full text: The determination of both the low- and high-LET components of ambient dose equivalent in mixed fields is possible with microdosimetric methods. With the multiple-event microdosimetric variance covariance method the sum of those components are directly obtained also in pulsed beams. However, if the value of each dose component is needed a more extended analysis is required. The use of a graphite walled proportional detector in combination with a tissue-equivalent proportional counter in combination with the variance covariance method was here investigated. MCNP simulations were carried out for relevant energies to investigate the photon and neutron responses of the two detectors. The combined graphite and TEPC system, the Sievert instrument, was used for measurements at IRSN, Cadarache, in the workplace calibration fields of CANEL+, SIGMA, a Cf-252 and a moderated Cf(D 2 O,Cd) radiation field. The response of the instrument in various monoenergetic neutron fields is also known from measurements at PTB. The instrument took part in the measurement campaigns in workplace fields in the nuclear industry organized within the EVIDOS contract. The results are analyzed and the method of using a graphite detector compared with alternative methods of analysis is discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  17. Discrimination of the wall effect in a thin counter with micro-gap structure for neutron position sensing

    Energy Technology Data Exchange (ETDEWEB)

    Sakae, Takeji; Manabe, Tohru; Kitamura, Yasunori; Nohtomi, Akihiro [Kyushu Univ., Fukuoka (Japan); Sakamoto, Sigeyasu

    1996-07-01

    Simulation by the Monte Carlo method is applied to estimate the wall effect in a thermal neutron counter having a new function for discriminating the effect. The counter is designed to have paralleled electrodes with micro-gap structure. A resistive anode is used for position sensing on the center of a set of the three electrode. The structure can be made by simple arrangement of anode and cathode wires on an insulator plane. The calculation shows discrimination of the wall effect can be achieved by coincident counting of two or three elements included in the counter. By using the coincident counting, the thickness of the neutron counter can be made into 1 mm with the information of the total energy created in the neutron detection. (author)

  18. Neutronic analysis of JET external neutron monitor response

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom (United Kingdom); Batistoni, Paola [ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden)

    2016-11-01

    Highlights: • We model JET tokamak containing JET remote handling system. • We investigate effect of remote handling system on external neutron monitor response. • Remote handling system correction factors are calculated. • Integral correction factors are relatively small, i.e up to 8%. - Abstract: The power output of fusion devices is measured in terms of the neutron yield which relates directly to the fusion yield. JET made a transition from Carbon wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) during 2010–11. Absolutely calibrated measurement of the neutron yield by JET neutron monitors was ensured by direct measurements using a calibrated {sup 252}Cf neutron source (NS) deployed by the in-vessel remote handling system (RHS) inside the JET vacuum vessel. Neutronic calculations were required in order to understand the neutron transport from the source in the vacuum vessel to the fission chamber detectors mounted outside the vessel on the transformer limbs of the tokamak. We developed a simplified computational model of JET and the JET RHS in Monte Carlo neutron transport code MCNP and analyzed the paths and structures through which neutrons reach the detectors and the effect of the JET RHS on the neutron monitor response. In addition we performed several sensitivity studies of the effect of substantial massive structures blocking the ports on the external neutron monitor response. As the simplified model provided a qualitative picture of the process only, some calculations were repeated using a more detailed full 3D model of the JET tokamak.

  19. Analysis of Induced Gamma Activation by D-T Neutrons in Selected Fusion Reactor Relevant Materials with EAF-2010

    Directory of Open Access Journals (Sweden)

    Klix Axel

    2016-01-01

    Full Text Available Samples of lanthanum, erbium and titanium which are constituents of structural materials, insulating coatings and tritium breeder for blankets of fusion reactor designs have been irradiated in a fusion peak neutron field. The induced gamma activities were measured and the results were used to check calculations with the European activation system EASY-2010. Good agreement for the prediction of major contributors to the contact dose rate of the materials was found, but for minor contributors the calculation deviated up to 50%.

  20. Comparisons of calculated and measured spectral distributions of neutrons from a 14-MeV neutron source inside the Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Barnes, J.M.; Alsmiller, R.G. Jr.; Emmett, M.B.; Drischler, J.D.

    1985-12-01

    A recent paper presented neutron spectral distributions (energy greater than or equal to0.91 MeV) measured at various locations around the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The neutron source for the series of measurements was a small D-T generator placed at various positions in the TFTR vacuum chamber. In the present paper the results of neutron transport calculations are presented and compared with these experimental data. The calculations were carried out using Monte Carlo methods and a very detailed model of the TFTR and the TFTR test cell. The calculated and experimental fluences per unit energy are compared in absolute units and are found to be in substantial agreement for five different combinations of source and detector positions

  1. Neutronic design and analysis on dual-cooled waste transmutation blanket for the fusion driven sub-critical system

    International Nuclear Information System (INIS)

    Zheng Shanliang; Wu Yican; Gao Chunjing; Xu Dezheng; Li Jingjing; Zhu Xiaoxiang

    2004-01-01

    Neutronics design and analysis of dual-cooled multi-functional waste transmutation blanket (DWTB) for the fusion driven sub-critical system (FDS) are performed to ensure the system be able to meet the requirements of fuel-sufficiency and more waste transmutation ratio with low initial loading fuel inventory, which is based on 1-D burn-up calculations with home-developed code Visual BUS and the multi-group (175 neutron groups-42 Gamma groups coupled) data library HENDL1.0/MG (Hybrid Evaluated Nuclear Data Library). (authors)

  2. Feasibility study of the water Cherenkov detector as a D-T fusion power monitor in the system using neutron activation of flowing water. First experimental phase

    International Nuclear Information System (INIS)

    Verzilov, Yury M.; Ochiai, Kentaro; Nishitani, Takeo

    2003-09-01

    The technique of monitoring D-T neutrons using water flow is based on the reaction of the 16 O(n, p) 16 N. In order to significantly improve the D-T neutron monitoring system in the ITER reactor in comparison with the system that uses a γ-ray scintillation detector, a new approach was proposed. The basic idea of this approach is to utilize the Cherenkov light, produced by energetic β-particles from 16 N in water near the first wall of the fusion reactor, and then deliver the light by the optical fiber to the remote light detector. The proof of the principle experiment is divided into two phases. The main idea of the first experimental phase is to examine Cherenkov light measurements using a remotely located water and light detector. During the second phase the water radiator will be placed next to the neutron source, then the Cherenkov light will be transferred by an optical fiber to the remotely located light detector. For the purpose of the first experimental phase, a water Cherenkov detector was installed in the shielded measurement room. A closed water loop, with circulating water, was used to transport 16 N from the D-T source to the Cherenkov detector. The experiment was carried out at FNS/JAERI, with the accelerator set to a direct current mode, the source neutron yield around 2 x 10 11 n/s, and the water flowage approximately 2 m/s. The registered Cherenkov signal was identified as the light produced by β-particles from 16 N using the time decay and the energy spectra data. According to the present study, the water Cherenkov detector is very effective for measurements of the 16 N activity, due to high counting efficiency, absence of the scintillation detector and simplicity of the method. (author)

  3. Production and use of Li(d,n) neutrons for simulation of radiation effects in fusion reactors

    International Nuclear Information System (INIS)

    Goland, A.N.; Gurinsky, D.H.; Hendrie, J.; Kukkonen, J.; Sheehan, T.; Snead, C.L. Jr.

    1975-01-01

    In the Brookhaven Accelerator-Based Neutron Generator 1.5-cm thick x 12-cm wide films of lithium flowing at the velocity of approximately 10 m sec -1 will be the targets for 30-MeV D + and D - beams 1-cm high and 10-cm wide. At this energy a beam of energetic neutrons is emitted mainly in the forward direction (theta less than or equal to 20 0 ) as a result of the Li(d,n) breakup reaction. Measurements of the neutron flux and spectrum as a function of incident deuteron energy and emission angle theta(theta less than or equal to 20 0 ) indicate that the yield increases approximately linearly with increasing deuteron energy from 25 MeV to at least 35 MeV, and that the mean energy of the neutrons (theta = 0 0 ) is about 0.4 of the incident deuteron energies between 25 and 35 MeV. The most probable neutron energy in the forward-directed (theta = 0 0 ) spectrum is also about 0.4 of the deuteron energy over this range. For a 30-MeV beam, the full width at half maximum of the neutron spectrum is 11.8 MeV (theta = 0 0 ), and the mean neutron energy is 13 MeV. Pertinent radiation-damage parameters were calculated for various materials exposed to this neutron spectrum. In Nb, for example, the helium production rate and the displacement rate simulate the values anticipated in a D-T fusion reactor spectrum of comparable flux. Furthermore, the primary-recoil-atom energy distributions produced by Li(d,n) neutrons in Al, Nb, and Au are similar to those produced by 14-MeV neutrons. (U.S.)

  4. Investigation of cascade-typed falling liquid film flow along first wall of laser-fusion reactor

    International Nuclear Information System (INIS)

    Kunugi, Tomoaki; Nakai, Tadakatsu; Kawara, Zensaku

    2007-01-01

    To protect from high energy/particle fluxes caused by nuclear fusion reaction such as extremely high heat flux, X rays, Alpha particles and fuel debris to a first wall of an inertia fusion reactor, a ''cascade-typed'' falling liquid film flow is proposed as the ''liquid wall'' concept which is one of the reactor chamber cooling and wall protection schemes: the reactor chamber can protect by using a liquid metal film flow (such as Li 17 Pb 83 ) over the wall. In order to investigate the feasibility of this concept, we conducted the numerical analyses by using the commercial code (STREAM: unsteady three-dimensional general purpose thermofluid code) and also conducted the flow visualization experiments. The numerical results suggested that the cascade structure design should be improved, so that we redesigned the cascade-typed first wall and performed the flow visualization as a POP (proof-of-principle) experiment. In the numerical analyses, the water is used as the working liquid and an acrylic plate as the wall. These selections are based on two reasons: (1) from the non-dimensional analysis approach, the Weber number (We=ru 2 d/s: r is density, u is velocity, d is film thickness, s is surface tension coefficient) should be the same between the design (Li 17 Pb 83 flow) and the model experiment (water flow) because of the free-surface instability, (2) the SiC/SiC composite would be used as the wall material, so that the wall may have the less wettability: the acrylic plate has the similar feature. The redesigned cascade-typed first wall for one step (30 cm height corresponding to 4 Hz laser duration) consists of a liquid tank having a free-surface for keeping the constant waterhead located at the backside of the first wall, and connects to a slit which is composed of two plates: one plate is the first wall, and the other is maintaining the liquid level. This design solved the trouble of the previous design. The test section for the flow visualization has the same

  5. Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices

    NARCIS (Netherlands)

    van Eden, G.G.; Kvon, V.; Van De Sanden, M.C.M.; Morgan, T.W.

    2017-01-01

    Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically

  6. AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

  7. Plasma induced material defects and threshold values for thermal loads in high temperature resistant alloys and in refractory metals for first wall application in fusion reactors

    International Nuclear Information System (INIS)

    Bolt, H.; Hoven, H.; Kny, E.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

    1986-10-01

    Materials for the application in the first wall of fusion reactors of the tokamak type are subjected to pulsed heat fluxes which range from some 0.5 MW m -2 to 10 MW m -2 during normal plasma operation, and which can exceed 1000 MW m -2 during total plasma disruptions. The structural defects and material fatigue caused by this types of plasma wall interaction are investigated and the results are plotted in threshold loading curves. Additionally, the results are, as far as possible, compared with quantitative, theoretical calculations. These procedures allow a semiquantitative evaluation of the applicability of the mentioned metals in the first wall of fusion reactors. (orig.) [de

  8. Neutron fluence at the reactor pressure vessel wall - a comparison of French and German procedures and strategies in PWRs

    International Nuclear Information System (INIS)

    Tricot, N.; Jendrich, U.

    2003-01-01

    While the neutrons within the core may take part in the chain reaction, those neutrons emitted from the core are basically lost for the energy production. This 'neutron leakage' represents a loss of fuel efficiency and causes neutron embrittlement of the reactor pressure vessel (RPV) wall. The latter raises safety concerns, needs to be monitored closely and may necessitate mitigating measures. There are different strategies to deal with these two undesirable effects: The neutron emission may be reduced to some extent all around the core or just at the 'hot spots' of RPV embrittlement by tailored core loading patterns. A higher absorption rate of neutrons may also be achieved by a larger water gap between the core and the RPV. In this paper the inter-relations between the distribution of neutron flux, core geometry, core loading strategy, RPV embrittlement and its surveillance are discussed at first. Then the different strategies followed by the German and French operators are described. Finally the conclusions will highlight the communalities and differences between these strategies as different approaches to the same problem of safety as well as economy. (authors)

  9. Critical survey of the neutron-induced creep behaviour of steel alloys for the fusion reactor materials programme

    International Nuclear Information System (INIS)

    Hausen, H.

    1985-01-01

    The differences between the irradiation environment of a fission reactor and that of a fusion reactor are respectively described in relation to the radiation damage found and expected in the two types of nuclear reactor. It is shown that the microstructure developing for instance in stainless steel alloys is almost invariant to whether the production rate of helium is high or low. The finding is valid up to neutron doses corresponding to about 60 dpa. For this reason, irradiation creep data obtained in fission reactors may be used, with caution, for predicting creep behaviour in fusion reactors.It was further recognized that irradiation creep performed with high energy particles from an accelerator, yields results which are comparable to those obtained in fission reactors. For this reason, simulation creep experiments are found to be valuable for the development of irradiation creep resistant materials using, for example, high energy electrons or protons. Such kind of experiments are performed in many laboratories. For irradiation doses larger than 60 dpa, predictions with respect to creep rates in fission and fusion reactors are difficult. In end-of-life tests, which concern swelling, ductility, tensile properties, rupture, fatigue and embrittlement, the presence of helium, due to its production rate being much higher in most materials exposed to 14 MeV neutrons than to fission neutrons, may be of great importance

  10. Coupled-channels analyses for 9,11Li + 208Pb fusion reactions with multi-neutron transfer couplings

    Science.gov (United States)

    Choi, Ki-Seok; Cheoun, Myung-Ki; So, W. Y.; Hagino, K.; Kim, K. S.

    2018-05-01

    We discuss the role of two-neutron transfer processes in the fusion reaction of the 9,11Li + 208Pb systems. We first analyze the 9Li + 208Pb reaction by taking into account the coupling to the 7Li + 210Pb channel. To this end, we assume that two neutrons are directly transferred to a single effective channel in 210Pb and solve the coupled-channels equations with the two channels. By adjusting the coupling strength and the effective Q-value, we successfully reproduce the experimental fusion cross sections for this system. We then analyze the 11Li + 208Pb reaction in a similar manner, that is, by taking into account three effective channels with 11Li + 208Pb, 9Li + 210Pb, and 7Li + 212Pb partitions. In order to take into account the halo structure of the 11Li nucleus, we construct the potential between 11Li and 208Pb with a double folding procedure, while we employ a Woods-Saxon type potential with the global Akyüz-Winther parameters for the other channels. Our calculation indicates that the multiple two-neutron transfer process plays a crucial role in the 11Li + 208Pb fusion reaction at energies around the Coulomb barrier.

  11. Monte Carlo simulation of high-flux 14 MeV neutron source based on muon catalyzed fusion using a high-power 50 MW deuteron beam

    Energy Technology Data Exchange (ETDEWEB)

    Vecchi, M [ENEA, Bologna (Italy); Karmanov, F I [Inst. of Nuclear Power Engineering, Obninsk (Russian Federation); Latysheva, L N; Pshenichnov, I A [Russian Academy of Sciences, Moscow (Russian Federation). Inst. for Nuclear Research

    1997-12-31

    The results Monte Carlo simulations of an intense neutron source based on muon catalyzed fusion process are presented. A deuteron beam is directed onto a cylindrical carbon target, located in vacuum converter chamber with a strong solenoidal magnetic field. The produced pions and muons which originate from pion decay are guided along magnetic field to a DT-synthesizer. Pion production in the primary target is simulated by means of Intranuclear and Internuclear cascade codes developed in INR, Moscow, while pion and muon transport process is studied by using a Monte Carlo code originated at CERN. The main purpose of the work is to calculate the pion and muon utilization efficiency taking into account the pion absorption in the primary target as well as all other losses of pions and muons in the converter and DT-cell walls. Preliminary estimations demonstrate the possibility to reach the level of 1014 n/s/cm{sup 2} for the neutron flux. (J.U.). 3 tabs., 4 figs., 8 refs.

  12. Moderator design studies for a new neutron reference source based on the D–T fusion reaction

    International Nuclear Information System (INIS)

    Mozhayev, Andrey V.; Piper, Roman K.; Rathbone, Bruce A.; McDonald, Joseph C.

    2016-01-01

    The radioactive isotope Californium-252 ( 252 Cf) is relied upon internationally as a neutron calibration source for ionizing radiation dosimetry because of its high specific activity. The source may be placed within a heavy-water (D 2 O) moderating sphere to produce a softened spectrum representative of neutron fields common to commercial nuclear power plant environments, among others. Due to termination of the U.S. Department of Energy loan/lease program in 2012, the expense of obtaining 252 Cf sources has undergone a significant increase, rendering high output sources largely unattainable. On the other hand, the use of neutron generators in research and industry applications has increased dramatically in recent years. Neutron generators based on deuteriumtritium (D–T) fusion reaction provide high neutron fluence rates and, therefore, could possibly be used as a replacement for 252 Cf. To be viable, the 14 MeV D–T output spectrum must be significantly moderated to approximate common workplace environments. This paper presents the results of an effort to select appropriate moderating materials and design a configuration to reshape the primary neutron field toward a spectrum approaching that from a nuclear power plant workplace. A series of Monte-Carlo (MCNP) simulations of single layer high- and low-Z materials are used to identify initial candidate moderators. Candidates are refined through a similar series of simulations involving combinations of 2–5 different materials. The simulated energy distribution using these candidate moderators are rated in comparison to a target spectrum. Other properties, such as fluence preservation and/or enhancement, prompt gamma production and other characteristics are also considered. - Highlights: • D–T generator neutron calibration field replacement for D 2 O-moderated 252 Cf. • Determination of representative nuclear power plant workplace neutron spectrum. • Simulations to assess moderating materials to soften 14

  13. A verification scenario of nuclear plus interference scattering effects using neutron incident angle distribution to the wall in beam-injected deuterium plasmas

    International Nuclear Information System (INIS)

    Sugiyama, Shota; Matsuura, Hideaki; Uchiyama, Daisuke; Sawada, Daisuke; Watanabe, Tsuguhiro; Goto, Takuya; Mitarai, Osamu

    2015-01-01

    A verification scenario of knock-on tail formation in the deuteron distribution function due to nuclear plus interference scattering is presented by observing the incident angle distribution of neutrons in a vacuum vessel. Assuming a knock-on tail created in a "3He-beam-injected deuterium plasma, the incident angle distribution and energy spectra of the neutrons produced by fusion reactions between 1-MeV and thermal deuterons are evaluated. The relation between the neutron incident angle to the vacuum vessel and neutron energy is examined in the case of anisotropic neutron emission due to knock-on tail formation in neutral-beam-injected plasmas. (author)

  14. Neutronic Parametric Study on a Conceptual Design for a Transmutation Fusion Blanket

    International Nuclear Information System (INIS)

    Tariq Siddique, M.; Kim, Myung Hyun

    2011-01-01

    Fusion energy may be the one of options of future energy. In all over the world, researchers are putting their efforts for its commercial and economical availability. Fusion-fission hybrid reactors have been studied for various applications in China. First milestone of fusion energy is expected to be the fusion fission hybrid reactors. In fusion-fission hybrid reactor the blanket design is of second prime importance after fusion source. In this study conceptual design of a fusion blanket is initiated for calculation of tritium production, transmutation of minor actinides (MA) and fission products (FP) and energy multiplication calculations

  15. Neutronic design of pulse operation simulating device for in-pile functional test of fusion blanket by MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Nagao, Yoshiharu; Nakamichi, Masaru; Kawamura, Hiroshi [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan)

    2000-03-01

    The pulse operation of a fusion reactor can be simulated in a fission reactor by controlling the neutron flux entering a test section by using a rotating 'hollow cylinder with window' made of hafnium. The rotating cylinder is installed between the test section and the fixed outer neutron absorber cylinder and is also made of hafnium with an opening in the direction to the core center. For gathering engineering data for the tritium breeding blanket such as characteristics of temperature change, tritium release and recovery, etc., it is desirable that the ratio of minimum to maximum thermal neutron fluxes is greater than 1:10. Design calculations were performed for the test assembly which considered local neutronic effects and the mechanical constraints of the device. From the results of these calculations, the ratio of minimum to maximum thermal neutron flux under irradiation would be about 1:10 using a pulse operation simulating device which has a thickness of 6.5 mm and a 150deg window angle for the rotating hollow cylinder and 5.0 mm in thickness of fixed neutron absorber. (author)

  16. Nondestructive Inspection System for Special Nuclear Material Using Inertial Electrostatic Confinement Fusion Neutrons and Laser Compton Scattering Gamma-Rays

    Science.gov (United States)

    Ohgaki, H.; Daito, I.; Zen, H.; Kii, T.; Masuda, K.; Misawa, T.; Hajima, R.; Hayakawa, T.; Shizuma, T.; Kando, M.; Fujimoto, S.

    2017-07-01

    A Neutron/Gamma-ray combined inspection system for hidden special nuclear materials (SNMs) in cargo containers has been developed under a program of Japan Science and Technology Agency in Japan. This inspection system consists of an active neutron-detection system for fast screening and a laser Compton backscattering gamma-ray source in coupling with nuclear resonance fluorescence (NRF) method for precise inspection. The inertial electrostatic confinement fusion device has been adopted as a neutron source and two neutron-detection methods, delayed neutron noise analysis method and high-energy neutron-detection method, have been developed to realize the fast screening system. The prototype system has been constructed and tested in the Reactor Research Institute, Kyoto University. For the generation of the laser Compton backscattering gamma-ray beam, a race track microtron accelerator has been used to reduce the size of the system. For the NRF measurement, an array of LaBr3(Ce) scintillation detectors has been adopted to realize a low-cost detection system. The prototype of the gamma-ray system has been demonstrated in the Kansai Photon Science Institute, National Institutes for Quantum and Radiological Science and Technology. By using numerical simulations based on the data taken from these prototype systems and the inspection-flow, the system designed by this program can detect 1 kg of highly enriched 235U (HEU) hidden in an empty 20-ft container within several minutes.

  17. European Fusion Programme. ITER task T23: Beryllium characterisation. Progress report. Tensile tests on neutron irradiated and reference beryllium

    International Nuclear Information System (INIS)

    Moons, F.

    1996-02-01

    As part of the European Technology Fusion Programme, the irradiation embrittlement characteristics of the more ductile and isotopic grades of beryllium manufactured by Brush Wellman has been investigated using modern powder production and consolidation techniques . This study was initiated in support of the development and evaluation of beryllium as a neutron multiplier for the solid breeder blanket design concepts proposed for a DEMO fusion power reactor. Four different species of beryllium: S-200 F (vacuum hot pressed, 1.2 wt% BeO), S-200FH (hot isostatic pressed, 0.9 wt% BeO), S-65 (vacuum hot pressed, 0.6 wt% BeO), S-65H (hot isostatic pressed, 0.5 wt% BeO) have been compared. Three batches of the beryllium have been investigated, a neutron batch, a thermal control batch and a reference batch. Neutron irradiation has been performed at temperatures between 175 and 605 degrees Celsius up to a neutron fluence of 2.1 10 25 n.m -2 (E> 1 MeV) or 750 appm He. The results of the tensile tests are summarized

  18. Neutronics shielding analysis of the last mirror-beam duct system for a laser fusion power reactor

    International Nuclear Information System (INIS)

    Ragheb, M.M.H.; Klein, A.C.

    1981-01-01

    A Monte Carlo three-dimensional neutronics analysis for the last mirror-beam duct system for the SOLASE conceptual laser-driven fusion power reactor design is presented. Detailed geometric configurations including the reactor cavity, the two last mirrors, and the three-section two-right-angle bends duct are modeled. Measurements are given of the dimensions and compositions of the reactor components, and of neutron scalar fluxes, spatial dependencies and neutron volumetric heating rates for the cases of aluminum or Boral as laser beam duct liners, and ordinary concrete or lead mortar as shield material. A three-dimensional modeling of laser-driven reactor penetrations is employed. The particle leakage is found to be excessively high for the configuration of the conceptual design considered and the advantages and disadvantages of various solutions, such as the use of Boral as a duct liner and the use of lead mortar instead of ordinary concrete as a shield material, are considered

  19. Influence of differences in the proton and neutron distributions on nuclear fusion and fission; Infuence de la difference entre les distributions de protons et de neutrons dans le noyau sur les processus de fusion et de fission

    Energy Technology Data Exchange (ETDEWEB)

    Dobrowolski, A

    2006-04-15

    This thesis work is centred on some essential ingredients of a theoretical description of the reaction dynamics of the nuclear fusion and fission process, such as the interaction potential between projectile and target nuclei for fusion and the deformation energy landscape in a multidimensional space for the fission process. We have in particular evaluated the importance of the difference between the neutron and proton density distributions on these 2 processes. The fusion potential between the two interacting nuclei is obtained through the nucleon densities, determined in a self-consistent way through semiclassical density variational calculations for a given effective nucleon-nucleon effective interaction of the Skyrme type. These fusion barriers can then be used in a Langevin formalism to evaluation fusion cross sections. For the fission process it turns out to be essential to allow for the large variety of shapes which appear between the nuclear ground state and the the scission configuration. We show that a shape parametrisation taking into account elongation, as well as possible neck formation, left-right asymmetry and non-axiality allows a precise description of this phenomena in the framework of the macroscopic-microscopic approach. We are thus able to enrich the expression of the liquid-drop type energy through a term which describes the variation of the nuclear energy due to a deformation difference between the proton and neutron distribution. The resulting reduction of the fission barriers is only of the order of one MeV but this can easily cause a change in the fission cross-section by an order of magnitude and thus plays a capital role for the stability of super-heavy of exotic nuclei. (author)

  20. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    International Nuclear Information System (INIS)

    Michael R, Kruzic

    2008-01-01

    centimeters squared (cm2) beta/gamma. Removable beta/gamma contamination levels seldom exceeded 1,000 dpm/100 cm2, but, in railroad trenches on the reactor pad containing soil on the concrete pad in front of the shield wall, the beta dose rates ranged up to 120 milli-roentgens per hour from radioactivity entrained in the soil. General area dose rates were less than 100 micro-roentgens per hour. Prior to demolition of the reactor shield wall, removable and fixed contaminated surfaces were decontaminated to the best extent possible, using traditional decontamination methods. Fifth, large sections of the remaining structures were demolished by mechanical and open-air controlled explosive demolition (CED). Mechanical demolition methods included the use of conventional demolition equipment for removal of three main buildings, an exhaust stack, and a mobile shed. The 5-foot (ft), 5-inch (in.) thick, neutron-activated reinforced concrete shield was demolished by CED, which had never been performed at the NTS

  1. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    Energy Technology Data Exchange (ETDEWEB)

    Michael R. Kruzic

    2008-06-01

    centimeters squared (cm2) beta/gamma. Removable beta/gamma contamination levels seldom exceeded 1,000 dpm/100 cm2, but, in railroad trenches on the reactor pad containing soil on the concrete pad in front of the shield wall, the beta dose rates ranged up to 120 milli-roentgens per hour from radioactivity entrained in the soil. General area dose rates were less than 100 micro-roentgens per hour. Prior to demolition of the reactor shield wall, removable and fixed contaminated surfaces were decontaminated to the best extent possible, using traditional decontamination methods. Fifth, large sections of the remaining structures were demolished by mechanical and open-air controlled explosive demolition (CED). Mechanical demolition methods included the use of conventional demolition equipment for removal of three main buildings, an exhaust stack, and a mobile shed. The 5-foot (ft), 5-inch (in.) thick, neutron-activated reinforced concrete shield was demolished by CED, which had never been performed at the NTS.

  2. Consultancy on the potential of fusion/fission sub-critical neutron systems for energy production and transmutation. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    The Workshop on Sub-critical Neutron Production held at the University of Maryland and the Eisenhower Institute on 11-13 October 2004 brought together members of fusion, fission and accelerator technical communities to discuss issues of spent fuel, nonproliferation, reactor safety and the use of neutrons for sub-critical operation of nuclear reactors. The Workshop strongly recommended that the fusion community work closely with other technical communities to ensure that a wider range of technical solutions is available to solve the spent fuel problem and to utilize the current actinide inventories. Participants of the Workshop recommended that a follow-on Workshop, possibly under the aegis of the IAEA, should be held in the first half of the year 2005. The Consultancy Meeting is the response to this recommendation. The objectives of the Consultancy meeting were to hold discussions on the role of fusion/fission systems in sub-critical operations of nuclear reactors. The participants agreed that development of innovative (fourth generation) fission reactors, advanced fuel cycle options, and disposition of existing spent nuclear fuel inventories in various Member Sates can significantly benefit from including sub-critical systems, which are driven by external neutron sources. Spallation neutrons produced by accelerators have been accepted in the past as the means of driving sub-critical reactors. The accelerator community deserves credit in pioneering this novel approach to reactor design. Progress in the design and operation of fusion devices now offers additional innovative means, broadening the range of sub-critical operations of fission reactors. Participants felt that fusion should participate with accelerators in providing a range of technical options in reactor design. Participants discussed concrete steps to set up a small fusion/fission system to demonstrate actinide burning in the laboratory and what advice should be given to the Agency on its role in

  3. Initial progress in the first wall, blanket, and shield Engineering Test Program for magnetically confined fusion-power reactors

    International Nuclear Information System (INIS)

    Herman, H.; Baker, C.C.; Maroni, V.A.

    1981-10-01

    The first wall/blanket/shield (FW/B/S) Engineering Test Program (ETP) progressed from the planning stage into implementation during July, 1981. The program, generic in nature, comprises four Test Program Elements (TPE's), the emphasis of which is on defining the performance parameters for the Fusion Engineering Device (FED) and the major fusion device to follow FED. These elements are: (1) nonnuclear thermal-hydraulic and thermomechanical testing of first wall and component facsimiles with emphasis on surface heat loads and heat transient (i.e., plasma disruption) effects; (2) nonnuclear and nuclear testing of FW/B/S components and assemblies with emphasis on bulk (nuclear) heating effects, integrated FW/B/S hydraulics and mechanics, blanket coolant system transients, and nuclear benchmarks; (3) FW/B/S electromagnetic and eddy current effects testing, including pulsed field penetration, torque and force restraint, electromagnetic materials, liquid metal MHD effects and the like; and (4) FW/B/S Assembly, Maintenance and Repair (AMR) studies focusing on generic AMR criteria, with the objective of preparing an AMR designers guidebook; also, development of rapid remote assembly/disassembly joint system technology, leak detection and remote handling methods

  4. Critical plasma-wall interaction issues for plasma-facing materials and components in near-term fusion devices

    International Nuclear Information System (INIS)

    Federici, G.; Coad, J.P.; Haasz, A.A.; Janeschitz, G.; Noda, N.; Philipps, V.; Roth, J.; Skinner, C.H.; Tivey, R.; Wu, C.H.

    2000-01-01

    The increase in pulse duration and cumulative run-time, together with the increase of the plasma energy content, will represent the largest changes in operation conditions in future fusion devices such as the International Thermonuclear Experimental Reactor (ITER) compared to today's experimental facilities. These will give rise to important plasma-physics effects and plasma-material interactions (PMIs) which are only partially observed and accessible in present-day experiments and will open new design, operation and safety issues. For the first time in fusion research, erosion and its consequences over many pulses (e.g., co-deposition and dust) may determine the operational schedule of a fusion device. This paper identifies the most critical issues arising from PMIs which represent key elements in the selection of materials, the design, and the optimisation of plasma-facing components (PFCs) for the first-wall and divertor. Significant advances in the knowledge base have been made recently, as part of the R and D supporting the engineering design activities (EDA) of ITER, and some of the most relevant data are reviewed here together with areas where further R and D work is urgently needed

  5. Elemental Characterization of Single-Wall Carbon Nanotube Certified Reference Material by Neutron and Prompt gamma Activation Analysis

    Czech Academy of Sciences Publication Activity Database

    Kučera, Jan; Bennett, J. W.; Oflaz, R.; Paul, R. L.; De Nadai Fernandes, E. A.; Kubešová, Marie; Bacchi, M. A.; Stopic, A. J.; Sturgeon, R. E.; Grinberg, P.

    2015-01-01

    Roč. 87, č. 7 (2015), s. 3699-3705 ISSN 0003-2700 R&D Projects: GA ČR(CZ) GBP108/12/G108; GA MŠk LM2011019 Institutional support: RVO:61389005 Keywords : Neutron Activation Analyses * nanotechnology * Carbon nanotubes * Chemical activation * Single-walled carbon nanotubes (SWCN) Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 5.886, year: 2015

  6. Implosion physics, alternative targets design and neutron effects on inertial fusion systems

    International Nuclear Information System (INIS)

    Velarde, G.; Martinez-Val, J.M.; Perlado, J.M.

    2001-01-01

    A new radiation transport code has been coupled with an existing multimaterial fluidynamics code using Adaptive Mesh Refinement (AMR) and its testing is presented, solving ray effect and shadow problems in SN classical methods. Important advances in atomic physics, opacity calculations and NLTE calculations, participating in significant experiments (LULI/France), have been obtained. Our new 1D target simulation model allows considering the effect of inverse Compton scattering in DT x targets (x<3%) working in a catalytic regime, showing the effectiveness of such tritium-less targets. Neutron activation of all natural elements in IFE reactors for waste management and that of target debris in NIF-type facilities have been completed. Pulse activation in structural walls is presented with a new modeling. Tritium atmospheric dispersion results indicate large uncertainties in environmental responses and needs to treat the two chemical forms. We recognise recombination barriers (metastable defects) and compute first systematic high-energy displacement cascade analysis in SiC, and radiation damage pulses by atomistic models in metals. Using Molecular Dynamics we explain the experimental evidence of low-temperature amorphization by damage accumulation in SiC. (author)

  7. Fast rigorous numerical method for the solution of the anisotropic neutron transport problem and the NITRAN system for fusion neutronics application. Pt. 1

    International Nuclear Information System (INIS)

    Takahashi, A.; Rusch, D.

    1979-07-01

    Some recent neutronics experiments for fusion reactor blankets show that the precise treatment of anisotropic secondary emissions for all types of neutron scattering is needed for neutron transport calculations. In the present work new rigorous methods, i.e. based on non-approximative microscopic neutron balance equations, are applied to treat the anisotropic collision source term in transport equations. The collision source calculation is free from approximations except for the discretization of energy, angle and space variables and includes the rigorous treatment of nonelastic collisions, as far as nuclear data are given. Two methods are presented: first the Ii-method, which relies on existing nuclear data files and then, as an ultimate goal, the I*-method, which aims at the use of future double-differential cross section data, but which is also applicable to the present single-differential data basis to allow a smooth transition to the new data type. An application of the Ii-method is given in the code system NITRAN which employs the Ssub(N)-method to solve the transport equations. Both rigorous methods, the Ii- and the I*-method, are applicable to all radiation transport problems and they can be used also in the Monte-Carlo-method to solve the transport problem. (orig./RW) [de

  8. Simultaneous neutron and x-ray imaging of inertial confinement fusion experiments along a single line of sight at Omega

    Energy Technology Data Exchange (ETDEWEB)

    Danly, C. R.; Day, T. H.; Herrmann, H.; Kim, Y. H.; Martinez, J. I.; Merrill, F. E.; Schmidt, D. W.; Simpson, R. A.; Volegov, P. L.; Wilde, C. H. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); Fittinghoff, D. N.; Izumi, N. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2015-04-15

    Neutron and x-ray imaging provide critical information about the geometry and hydrodynamics of inertial confinement fusion implosions. However, existing diagnostics at Omega and the National Ignition Facility (NIF) cannot produce images in both neutrons and x-rays along the same line of sight. This leads to difficulty comparing these images, which capture different parts of the plasma geometry, for the asymmetric implosions seen in present experiments. Further, even when opposing port neutron and x-ray images are available, they use different detectors and cannot provide positive information about the relative positions of the neutron and x-ray sources. A technique has been demonstrated on implosions at Omega that can capture x-ray images along the same line of sight as the neutron images. The technique is described, and data from a set of experiments are presented, along with a discussion of techniques for coregistration of the various images. It is concluded that the technique is viable and could provide valuable information if implemented on NIF in the near future.

  9. Surface roughness effect on ultracold neutron interaction with a wall and implications for computer simulations

    OpenAIRE

    Steyerl, A.; Malik, S. S.; Desai, A. M.; Kaufman, C.

    2009-01-01

    We review the diffuse scattering and the loss coefficient in ultracold neutron reflection from slightly rough surfaces, report a surprising reduction in loss coefficient due to roughness, and discuss the possibility of transition from quantum treatment to ray optics. The results are used in a computer simulation of neutron storage in a recent neutron lifetime experiment that re-ported a large discrepancy of neutron lifetime with the current particle data value. Our partial re-analysis suggest...

  10. Phase IIC experiments of the USDOE/JAERI collaborative program on fusion blanket neutronics

    International Nuclear Information System (INIS)

    Youssef, M.Z.

    1992-12-01

    Effort in Phase IIC of the US/JAERI Collaborative Program on Fusion Neutronics was focused on performing integral experiments and post analyses on blankets that include the actual heterogeneities found in several blanket designs. Two geometrical arrangements were considered for the blanket assembly, namely multi-layers of Li 2 O and beryllium in an edge-on, horizontally alternating configuration for a front depth of 30 cm, followed by the Li 2 O breeding zone (Be edge-on, BEO, experiment), and vertical water coolant channels arrangement (WCC experiment). The objectives are to examine the accuracy of predicting tritium production. In the BEO system, it was shown that, with the zonal method to measure tritium production from natural lithium (Tn), the calculated-to-measured values (C/E) are 0.95-1.05 (JAERI) and 0.98-0.9 (U.S.), which is consistent with the results obtained in other Phases of the Program (Phases IIA and IIb)). In the WCC experiment, there is a noticeable change in C/E values for T 6 near the coolant channels where steep gradients in T 6 production are observed. The C/E values obtained with the Li-foil detectors are on the average closer to unity than those obtained by the Li-glass method. As for T 7 , the values obtained by NE213 method are within ±15% in JAERI's calculations, but larger values (∼20-25%) are obtained in the U.S. calculations due to the differences of cross-sections data files. Around heterogeneities, the prediction accuracy for T 7 is better than for T 6 . (J.P.N.)

  11. Neutron bombardment of single wall carbon nanohorn (SWCNH). DSC determination of the stored Wigner-Szilard energy

    International Nuclear Information System (INIS)

    Franco Cataldo; Susana Iglesias-Groth; Yaser Hafez; Giancarlo Angelini

    2014-01-01

    Single wall carbon nanohorn (SWCNH) were neutron-bombarded to a dose of 3.28 × 10 16 n/cm 2 . The Wigner or stored energy was determined by a differential scanning calorimeter and was found 5.49 J/g, 50 times higher than the Wigner energy measured on graphite flakes treated at the same neutron dose. The activation energy for the thermal annealing of the accumulated radiation damage in SWCNH was determined in the range 6.3-6.6 eV against a typical activation energy for the annealing of the radiation-damaged graphite which is in the range of 1.4-1.5 eV. Furthermore the stored energy in neutron-damaged SWCNH is released at 400-430 deg C while the main peak in the neutron-damaged graphite occurs at 200-220 deg C. The radiation damaged SWCNH were examined with FT-IR spectroscopy showing the formation of acetylenic and aliphatic moieties suggesting the aromatic C=C breakdown caused by the neutron bombardment. (author)

  12. Mechanical properties of materials in fusion reactor first-wall and blanket systems

    Energy Technology Data Exchange (ETDEWEB)

    Bloom, E.E.

    1979-01-01

    With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n,..cap alpha..) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above Tm/2. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their melting temperatures.

  13. Mechanical properties of materials in fusion reactor first-wall and blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.

    1979-01-01

    With respect to the effects of irradiation on mechanical properties, the most significant difference between fast fission and fusion reactor spectra is the relatively large amount of helium produced by (n,α) transmutations in the latter. Relevant information on the effects of large amounts of helium (with concomitant displacement damage) comes from irradiation of alloys containing nickel in mixed spectrum reactors. At helium levels of interest for fusion reactor development, properties are degraded to unacceptable levels above Tm/2. Below this temperature, strength and ductility are retained and fractures remain transgranular. Importantly, the properties remain sensitive to composition and structure. A comparison of the response of bcc refractory alloys to that of stainless steel at equivalent damage levels shows the same general trends in properties with homologous temperature. The refractory alloys do offer potential for higher temperature applications because of their melting temperatures

  14. First wall/blanket/shield design and power conversion for the ARIES-IV tokamak fusion reactor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Conn, R.W.; Najmabadi, F.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10 MPa base pressure. The coolant flows poloidally in two loops, one inboard and one outboard. The coolant channels are circular tubes that form shells and are placed between two purge plates; the space between two adjacent tubes and the plate is purge gas flow area. The solid breeder is Li 2 O, and Be is used as neutron multiplier to ensure adequate TBR. Beryllium and Li 2 O are placed in between the adjacent tube shells. A computer code was developed to perform and optimize thermal-hydraulic design. Minimization of blanket thickness and the amount of Be, and the maximization of breeder zone thickness were done by iteration with neutronics. The gross thermal efficiency is 49%. The cost of electricity is 68 mills/kWh. The use of low activation SiC composite as the structural material, Li 2 O as the solid breeder, and avoidance of tungsten in the divertor has resulted in a good safety performance, and LSA rating of 1. Overall, SiC/He/Li 2 O ARIES-IV design is expected to have attractive economic and safety advantages

  15. Plasma sprayed TiC coatings for first wall protection in fusion devices

    International Nuclear Information System (INIS)

    Groot, P.; Laan, J.G. van der; Laas, L.; Mack, M.; Dvorak, M.

    1989-01-01

    For protection of plasma facing components in nuclear fusion devices thick titanium carbide coatings are being developed. Coatings have been produced by plasma spraying at atmospheric pressure (APS) and low pressure (LPPS) and analyzed with respect to microstructure and chemical composition. Thermo-mechanical evaluation has been performed by applying short pulse laser heat flux tests. The influence of coating thickness and porosity on the resistance to spalling by thermal shocks appears to be more important than aspects of chemical composition. (author)

  16. Effects of low-temperature fusion neutron irradiation on critical properties of a monofilament niobium-tin superconductor

    International Nuclear Information System (INIS)

    Guinan, M.W.; Van Konynenburg, R.A.; Mitchell, J.B.

    1984-01-01

    The objective of this work was to irradiate a Nb 3 Sn superconductor with 14.8 MeV neutrons at 4 K and measure critical current in transverse fields of up to 12 T, irradiating up to a fluence sufficient to decrease the critical current to below its initial value. Critical temperatures were also to be measured. The samples were to be kept near 4 K between the irradiation and the measurement of critical properties. This work is directed toward establishing an engineering design fluence limit for Nb 3 Sn when used in fusion reactor superconducting magnets

  17. Effects of low-temperature fusion neutron irradiation on critical properties of a monofilament niobium-tin superconductor

    Energy Technology Data Exchange (ETDEWEB)

    Guinan, M.W.; Van Konynenburg, R.A.; Mitchell, J.B.

    1984-03-22

    The objective of this work was to irradiate a Nb/sub 3/Sn superconductor with 14.8 MeV neutrons at 4 K and measure critical current in transverse fields of up to 12 T, irradiating up to a fluence sufficient to decrease the critical current to below its initial value. Critical temperatures were also to be measured. The samples were to be kept near 4 K between the irradiation and the measurement of critical properties. This work is directed toward establishing an engineering design fluence limit for Nb/sub 3/Sn when used in fusion reactor superconducting magnets.

  18. Neutron irradiation of V-Cr-Ti alloys in the BOR-60 fast reactor: Description of the fusion-1 experiment

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F. [Oak Ridge National Laboratory, TN (United States); Tsai, H.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    The FUSION-1 irradiation capsule was inserted in Row 5 of the BOR-60 fast reactor in June 1995. The capsule contains a collaborative RF/U.S. experiment to investigate the irradiation performance of V-Cr-Ti alloys in the temperature range 310 to 350{degrees}C. This report describes the capsule layout, specimen fabrication history, and the detailed test matrix for the U.S. specimens. A description of the operating history and neutronics will be presented in the next semiannual report.

  19. The thermal response of the first wall of a fusion reactor blanket to plasma disruptions

    International Nuclear Information System (INIS)

    Klippel, H.Th.

    1983-09-01

    Major plasma disruptions in Tokamak power reactors are potentially dangerous because high thermal overloading of the first wall may occur, resulting in melting and evaporation. The present uncertainties of the disruption characteristics, in particular the space and time dependence of the energy deposition, lead to a wide variation in the prospective surface energy loads. The thermal response of a first wall of aluminium, stainless steel and of graphite subjected to disruption energy loads up to 1000 J cm -2 has been analysed including the effects of melting and surface evaporation, vapour recondensation, vapour shielding, and the moving of the surface boundary caused by the evaporation. A special calculation model has been developed for this purpose. The main results are the following: by values of local transient energy depositions over 1500 J cm -2 bare stainless steel walls are damaged severely. Further calculations are needed to estimate the endurance limit of several candidate first wall materials. Applications of coatings on surfaces need special attention. For the reference INTOR disruption (approx. 100 J cm -2 ) evaporation is not significant. The effect of vapour shielding on evaporation has been found to be significant. The effect on melting is less pronounced. In a complete analysis the stability and dynamic behaviour of the melted layer under electromagnetic forces should be included. Also a reliable set of plasma disruption characteristics should be gathered

  20. Evaluation of copper alloys for fusion reactor divertor and first wall components

    DEFF Research Database (Denmark)

    Fabritsiev, S.A.; Zinkle, S.J.; Singh, B.N.

    1996-01-01

    This paper presents a critical analysis of the main factors of radiation damage limiting the possibility to use copper alloys in the ITER divertor and first wall structure. In copper alloys the most significant types of radiation damage in the proposed temperature-dose operation range are swellin...

  1. Experimental study of angular neutron flux spectra on a slab surface to assess nuclear data and calculational methods for a fusion reactor design

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1988-06-01

    This paper presents an experimental approach to interpret the results of integral experiments for fusion neutronics research. The measurement is described of the angular neutron flux on a restricted area of slab assemblies with D-T neutron bombardment by using the time-of-flight (TOF) method with an NE213 neutron detector over an energy range from 0.05 to 15 MeV. A two bias scheme was developed to obtain an accurate detection efficiency over a wide energy range. The detector-collimator response function was introduced to define the restricted surface area and to determine the effective measured area. A series of measurements of the angular neutron flux on slabs of fusion blanket materials, i.e., Be, C, and Li 2 O, as functions of neutron leaking angle and slab thickness have been performed to examine neutron transport characteristics in bulk materials. The calculational analyses of the experimental results have been also carried out by using Monte Carlo neutron transport codes, i.e., MORSE-DD and MCNP. The existing nuclear data files, i.e., JENDL-3PR1, -3PR2, ENDF/B-IV and -V were tested by comparing with the experimental results. From the comparisons, the data on C and 7 Li in the present files are fairly sufficient. Those on beryllium, however, is insufficient for the estimation of high threshold reactions such as tritium production in a fusion reactor blanket design. It is also found that the total and elastic cross sections are more important for accurate predictions of neutronic parameters at deep position. The comparisons between the measured and calculated results provide information to understand the results of the previous integral experiments for confirmation of accuracy of fusion reactor designs. (author)

  2. Resistive requirements for the vacuum wall of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Brooks, J.N.; Ehat, D.; Harkness, S.D.; Norem, J.; Stevens, H.; Turner, L.

    1978-01-01

    Most conceptual designs of tokamak power reactors have incorporated a ceramic insulator in the vacuum wall to make the wall electrically non-conducting. Such a material will have to be highly resistant to radiation damage at doses up to at least 10 MW-yr/m 2 while being compatible with a coolant and a first wall whose dimensions change due to thermal cycling and radiation damage. Thus there is considerable incentive to assess the consequences of eliminating the flux breaker from the design and having a conducting boundary instead. In this initial study the question of having a finite wall resistance has been examined in terms of its major implications on both the normal and abnormal operation of a tokamak reactor. This study has been conducted within the framework of the ANL-EPR-77 design although the results should provide some guidance for future reactors as well. The EPR design referred to is a 5 m major radius tokamak with an aspect ratio of 3.5, and with an equilibrium plasma current of 7.3 MA. The vacuum chamber is designed to accommodate a non-circular plasma with a height to width ratio of up to 1.65. The basic vacuum wall design is shown in Fig. 1. It is located about 0.4 M from the plasma boundary and has an irregular polygon shape made of sixteen sections, one per TF coil interval. Variations of this design having a range of resistance values have been used in the analysis

  3. Surface roughness effect on ultracold neutron interaction with a wall and implications for computer simulations

    International Nuclear Information System (INIS)

    Steyerl, A.; Malik, S. S.; Desai, A. M.; Kaufman, C.

    2010-01-01

    We review the diffuse scattering and the loss coefficient in ultracold neutron reflection from slightly rough surfaces, report a surprising reduction in loss coefficient due to roughness, and discuss the possibility of transition from quantum treatment to ray optics. The results are used in a computer simulation of neutron storage in a recent neutron lifetime experiment that reported a large discrepancy of neutron lifetime with the current particle data value. Our partial reanalysis suggests the possibility of systematic effects that were not included in this publication.

  4. Excitation of surface waves of ultracold neutrons on absorbing trap walls as anomalous loss factor

    International Nuclear Information System (INIS)

    Bokun, R.Ch.

    2006-01-01

    One analyzed probability of excitation of surface waves of ultracold neutrons in terms of a plane model consisting of three media: vacuum, a finite depth neutron absorbing substance layer and a neutron reflecting substrate. One demonstrated the absence of the mentioned surface waves in terms of the generally accepted model of two media: vacuum contiguous to the plane surface of a substance filled half-space. One pointed out the effect of the excited surface waves of ultracold neutrons on the increase of their anomalous losses in traps [ru

  5. Scientific report. Plasma-wall interaction studies related to fusion reactor materials

    International Nuclear Information System (INIS)

    Temmerman, G. De

    2006-01-01

    This scientific report summarises research done on erosion and deposition mechanisms affecting the optical reflectivity of potential materials for use in the mirrors used in fusion reactors. Work done in Juelich, Germany, at the Federal Institute of Technology in Lausanne, Switzerland, the JET laboratory in England and in Basle is discussed. Various tests made with the mirrors are described. Results obtained are presented in graphical and tabular form and commented on. The influence of various material choices on erosion and deposition mechanisms is discussed

  6. Advanced management of pipe wall thinning based on prediction-monitor fusion

    International Nuclear Information System (INIS)

    Kojima, Fumio; Uchida, Shunsuke

    2012-01-01

    This article is concerned with pipe wall thinning management system by means of hybrid use of simulation and monitoring. First, the computer-aided simulation for predicting wear rate of piping system is developed based on elucidation of thinning mechanism such as flow-accelerated corrosion (FAC). The accurate prediction of wear rate allows us the useful information on region of interest of inspection. Secondly, several monitoring methods are considered in accordance with interest of inspection. Thirdly, probability of detection (POD) is considered for the reliability of inspection data. The final part of this article is devoted to how to improve safety performance under the hybrid use of predicting and monitoring on the proposed pipe wall management. (author)

  7. Impurity effects in neutron-irradiated simple oxides: Implications for fusion devices

    International Nuclear Information System (INIS)

    Gonzalez, R.; Chen, Y.; Caceres, D.; Vergara, I.

    2006-01-01

    Radiation damage induced by neutron irradiation was studied in undoped MgO crystals and in MgO doped with either iron, hydrogen or lithium impurities. The oxygen-vacancy concentration produced by irradiation increases with neutron fluence. The net production rates resulting from irradiations with 14.8 MeV neutrons are about twice those produced by fission neutrons. In nominally pure crystals, the oxygen-vacancy concentration incurred by the fission-neutron irradiation is higher in crystals with a larger number of inherent impurities (such as iron) due to trapping of interstitials by impurities. Suppression of these defects is observed in MgO:H crystals and attributed to migration of oxygen vacancies to microcavities filled with H 2 gas. In MgO:Li crystals irradiated with neutron fluences below 10 18 n/cm 2 , most of the oxygen vacancies are camouflaged as hydride ions. Nanoindentation experiments show that hardness increases with neutron fluence and is independent of the presence of lithium in the crystal. Comparison between a neutron-irradiated and a thermochemically reduced crystal containing similar concentrations of oxygen vacancies shows that 70% of the neutron-irradiation hardening is produced by interstitials, 30% by oxygen vacancies and a negligible amount by higher-order point defects

  8. Level crossings, excess times, and transient plasma–wall interactions in fusion plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Theodorsen, A., E-mail: audun.theodorsen@uit.no; Garcia, O. E., E-mail: odd.erik.garcia@uit.no [Department of Physics and Technology, UiT The Arctic University of Norway, N-9037 Tromsø (Norway)

    2016-04-15

    Based on a stochastic model for intermittent fluctuations in the boundary region of magnetically confined plasmas, an expression for the level crossing rate is derived from the joint distribution of the process and its derivative. From this the average time spent by the process above a certain threshold level is obtained. This provides novel predictions of plasma–wall interactions due to transient transport events associated with a radial motion of blob-like structures in the scrape-off layer.

  9. Temperature and displacement transients in inertial confinement fusion first-walls

    International Nuclear Information System (INIS)

    Hunter, T.O.; Kulcinski, G.L.

    1977-01-01

    A quasi-analytic general model is developed for determination of temperature response and displacement damage in materials exposed to bursts of thermonuclear radiations. Temperature response can be determined for any time or position. Materials are assessed, using the model, which might be employed for dry first walls, collectors, laser mirrors, or other exposed reactor components. The resulting magnitude and temporal distribution of temperature and displacement production show that effects on material micro-structure must be treated in a dynamic fashion

  10. An assessment of the evaporation and condensation phenomena of lithium during the operation of a Li(d,xn fusion relevant neutron source

    Directory of Open Access Journals (Sweden)

    J. Knaster

    2016-12-01

    Full Text Available The flowing lithium target of a Li(d,xn fusion relevant neutron source must evacuate the deuteron beam power and generate in a stable manner a flux of neutrons with a broad peak at 14 MeV capable to cause similar phenomena as would undergo the structural materials of plasma facing components of a DEMO like reactors. Whereas the physics of the beam-target interaction are understood and the stability of the lithium screen flowing at the nominal conditions of IFMIF (25 mm thick screen with +/–1 mm surface amplitudes flowing at 15 m/s and 523 K has been demonstrated, a conclusive assessment of the evaporation and condensation of lithium during operation was missing. First attempts to determine evaporation rates started by Hertz in 1882 and have since been subject of continuous efforts driven by its practical importance; however intense surface evaporation is essentially a non-equilibrium process with its inherent theoretical difficulties. Hertz-Knudsen-Langmuir (HKL equation with Schrage’s ‘accommodation factor’ η = 1.66 provide excellent agreement with experiments for weak evaporation under certain conditions, which are present during a Li(d,xn facility operation. An assessment of the impact under the known operational conditions for IFMIF (574 K and 10−3Pa on the free surface, with the sticking probability of 1 inherent to a hot lithium gas contained in room temperature steel walls, is carried out. An explanation of the main physical concepts to adequately place needed assumptions is included.

  11. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

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

  12. 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...

  13. ZZ DECNET-GENDF, Fusion Damage Library of 175 Neutron and 42 Photon VITAMIN-J Groups

    International Nuclear Information System (INIS)

    1997-01-01

    1 - Description of program or function: DECNET is a library for fusion damage computations of 175 neutron + 42 photon VITAMIN-J energy group with the standard weighting function: Maxwellian (at the temperature to which the material is referenced) + 1/E + Fission Spectrum + 1/E + Fusion Peak + 1/E; it includes neutron kerma and gamma-ray production data from radioactive nuclei at 3 temperatures with the same materials of ZZ-GEFF-2-GENDF (see below) from 1-H-1 to Bi-209, mostly taken from EFF-2 with some nuclides from JEF-2.2 - Ag-107, Ag-109, Cd, the 6 Hf isotopes and the 4 W isotopes; however the list of the materials disagrees with that of GEFF-2 in that all elemental nuclides have been split into the components isotopes to follow the respective decay chain and not all materials of GEFF-2 produces nuclei which disintegrate. The library has been produced by the DECKER code which has been developed for this purpose. The format of the library is GENDF. 2 - Method of solution: The library has been produced by the DECKER code developed at ENEA Bologna for this purpose. The code reads the nuclide(s) for which decay kerma and photon production are requested and looks for the necessary data on the RDD (Radioactive Decay Data) file from JEF-2.2

  14. Pulsed Operation of a Compact Fusion Neutron Source Using a High-Voltage Pulse Generator Developed for Landmine Detection

    International Nuclear Information System (INIS)

    Yamauchi, Kunihito; Watanabe, Masato; Okino, Akitoshi; Kohno, Toshiyuki; Hotta, Eiki; Yuura, Morimasa

    2005-01-01

    Preliminary experimental results of pulsed neutron source based on a discharge-type beam fusion called Inertial Electrostatic Confinement Fusion (IECF) for landmine detection are presented. In Japan, a research and development project for constructing an advanced anti-personnel landmine detection system by using IECF, which is effective not only for metal landmines but also for plastic ones, is now in progress. This project consists of some R and D topics, and one of them is R and D of a high-voltage pulse generator system specialized for landmine detection, which can be used in the severe environment such as that in the field in Afghanistan. Thus a prototype of the system for landmine detection was designed and fabricated in consideration of compactness, lightness, cooling performance, dustproof and robustness. By using this prototype pulse generator system, a conventional IECF device was operated as a preliminary experiment. As a result, it was confirmed that the suggested pulse generator system is suitable for landmine detection system, and the results follow the empirical law obtained by the previous experiments. The maximum neutron production rate of 2.0x10 8 n/s was obtained at a pulsed discharge of -51 kV, 7.3 A

  15. DETECTORS AND EXPERIMENTAL METHODS: Studies of a scintillator-bar detector for a neutron wall at an external target facility

    Science.gov (United States)

    Yu, Yu-Hong; Xu, Hua-Gen; Xu, Hu-Shan; Zhan, Wen-Long; Sun, Zhi-Yu; Guo, Zhong-Yan; Hu, Zheng-Guo; Wang, Jian-Song; Chen, Jun-Ling; Zheng, Chuan

    2009-07-01

    To achieve a better time resolution of a scintillator-bar detector for a neutron wall at the external target facility of HIRFL-CSR, we have carried out a detailed study of the photomultiplier, the wrapping material and the coupling media. The timing properties of a scintillator-bar detector have been studied in detail with cosmic rays using a high and low level signal coincidence. A time resolution of 80 ps has been achieved in the center of the scintillator-bar detector.

  16. Characterization of graded iron / tungsten layers for the first wall of fusion reactors; Charakterisierung gradierter Eisen/Wolfram-Schichten fuer die erste Wand von Fusionsreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Heuer, Simon

    2017-07-01

    The nuclear fusion has great potential to enable a CO{sub 2}-neutral energy supply of future generations. The technical utilization of this energy source has hitherto been a challenge. In particular, high thermal loads and neutron-induced damage lead to extreme demands on the choice of materials for plasma-facing components (PFCs). These are therefore, as currently understood, made from a tungsten protective layer which is joined to a structure of low activation ferritic-martensitic (LAFM) steel. Due to the discrete transition of material properties at the LAFM-W joining zone as well as thermal loads, macroscopic stresses and plastic strains arise here. A feasible way to reduce this is to implement an intermediate layer with graded LAFM / W ratio, a so-called functional graded material (FGM). In the present work, macro-stresses and strains in the first wall of the fusion reactor DEMO are examined and evaluated by means of a finite element simulation. In this framework model components with and without graded interlayer are taken into account and the advantage of a FGM is emphasized. Parameter studies serve as a constructive guideline for the structural implementation of FGMs and components of the first wall. In addition, the feasibility of four methods (magnetron sputtering, liquid phase infiltration, modified atmospheric plasma spraying and electrodischarge sintering) with respect to the fabrication of FGMs is being studied. The resulting layers are microstructurally, thermo-physically and mechanically examined in detail. Based on this characterization and the finite element simulation, their suitability as a graded layer in the first wall of DEMO is evaluated and finally compared with alternative joining systems that are currently being tested in the research environment. [German] Die Kernfusion besitzt grosses Potenzial eine CO{sub 2}-neutrale Energieversorgung zukuenftiger Generationen zu ermoeglichen. Dabei stellt die technische Nutzbarmachung dieser

  17. Reactor concepts for laser fusion

    International Nuclear Information System (INIS)

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

    1977-07-01

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

  18. Effects of magnetization on fusion product trapping and secondary neutron spectra

    International Nuclear Information System (INIS)

    Knapp, P. F.; Schmit, P. F.; Hansen, S. B.; Gomez, M. R.; Hahn, K. D.; Sinars, D. B.; Peterson, K. J.; Slutz, S. A.; Sefkow, A. B.; Awe, T. J.; Harding, E.; Jennings, C. A.; Desjarlais, M. P.; Chandler, G. A.; Cooper, G. W.; Cuneo, M. E.; Geissel, M.; Harvey-Thompson, A. J.; Porter, J. L.; Rochau, G. A.

    2015-01-01

    By magnetizing the fusion fuel in inertial confinement fusion (ICF) systems, the required stagnation pressure and density can be relaxed dramatically. This happens because the magnetic field insulates the hot fuel from the cold pusher and traps the charged fusion burn products. This trapping allows the burn products to deposit their energy in the fuel, facilitating plasma self-heating. Here, we report on a comprehensive theory of this trapping in a cylindrical DD plasma magnetized with a purely axial magnetic field. Using this theory, we are able to show that the secondary fusion reactions can be used to infer the magnetic field-radius product, BR, during fusion burn. This parameter, not ρR, is the primary confinement parameter in magnetized ICF. Using this method, we analyze data from recent Magnetized Liner Inertial Fusion experiments conducted on the Z machine at Sandia National Laboratories. We show that in these experiments BR ≈ 0.34(+0.14/−0.06) MG · cm, a ∼ 14× increase in BR from the initial value, and confirming that the DD-fusion tritons are magnetized at stagnation. This is the first experimental verification of charged burn product magnetization facilitated by compression of an initial seed magnetic flux

  19. Low cycle fatigue lifetime of HIP bonded Bi-metallic first wall structures of fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hatano, Toshihisa; Sato, Satoshi; Furuya, Kazuyuki; Kuroda, Toshimasa; Enoeda, Mikio; Takatsu, Hideyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Hashimoto, Toshiyuki; Kitamura, Kazunori

    1998-10-01

    A HIP bonded bi-metallic panel composed of a dispersion strengthened copper (DSCu) layer and type 316L stainless steel (SS316L) cooling pipes is the reference design of the ITER first wall. To examine the fatigue lifetime of the first wall panel under cyclic mechanical loads, low cycle fatigue tests of HIP bonded bi-metallic specimens made of SS316L and DSCu were conducted with the stress ratio of -1.0 and five nominal strain range conditions ranging from 0.2 to 1.0%. Elasto-plastic analysis has also been conducted to evaluate local strain ranges under the nominal strains applied. Initial cracks were observed at the inner surface of the SS316L cooling pipes for all of the specimens tested, which was confirmed by the elasto-plastic analysis that the maximum strains of the test specimens were developed at the same locations. It was found that the HIP bonded bi-metallic test specimens had a fatigue lifetime longer than that of the SS316L raw material obtained by round bar specimens. Similarly, the fatigue lifetime of the DSCu/SS316L HIP interface was also longer than the round bar test results for the HIP joints. From these results, it has been confirmed that the bi-metallic first wall panel with built-in cooling pipes made by HIP bonding has a sufficient fatigue lifetime in comparison with the raw fatigue data of the materials, which also suggests that the fatigue lifetime evaluation has an adequate margin against fracture if it follows the design fatigue curve based on the material fatigue data. (author)

  20. Radiation effects on structural ceramics in fusion

    International Nuclear Information System (INIS)

    Hopkins, G.R.; Price, R.J.; Trester, P.W.

    1986-01-01

    Ceramics are required to serve in a conventional role as electrical and thermal insulators and dielectrics in fusion power reactors. In addition, certain ceramic materials can play a unique structural role in fusion power reactors by virtue of their very low induced radioactivity from fusion neutron capture. The aspects of safety, long-term radioactive waste management, and personnel access for maintenance and repair can all be significantly improved by applying the low-activation ceramics to the structural materials of the first-wall and blanket regions of a fusion reactor. Achievement of long service life at high structural loads and thermal stresses on the materials exposed to high-radiation doses presents a critical challenge for fusion. In this paper, we discuss radiation effects on structural ceramics for fusion application