Mechanisms of hydrogen retention in metallic beryllium and beryllium oxide and properties of ion-induced beryllium nitride

International Nuclear Information System (INIS)

Oberkofler, Martin

2011-01-01

In the framework of this thesis laboratory experiments on atomically clean beryllium surfaces were performed. They aim at a basic understanding of the mechanisms occurring upon interaction of a fusion plasma with a beryllium first wall. The retention and the temperature dependent release of implanted deuterium ions are investigated. An atomistic description is developed through simulations and through the comparison with calculations based on density functional theory. The results of these investigations are compared to the behaviour of hydrogen upon implantation into thermally grown beryllium oxide layers. Furthermore, beryllium nitride is produced by implantation of nitrogen into metallic beryllium and its properties are investigated. The results are interpreted with regard to the use of beryllium in a fusion reactor. (orig.)

Analysis of influence of fast neutron fluence irradiated to Beryllium element of The RSG-GAS reactor

International Nuclear Information System (INIS)

Sri Kuntjoro

2010-01-01

Analysis of influence fast neutron fluence irradiated to the RSG-GAS beryllium reflector have been done. Methods of analysis was carried out by measuring fluxes neutron in beryllium element and block position that function as reflector.The calculation done for determination it is there any influence of neutron as long as beryllium in the core. Besides that, visualization done to make sure it there is any deformation at beryllium as effect of irradiation. Fluxes and fluences of beryllium element measurement result in 200 kW reactor power are 2.30E+07 n/cm 2 .sec and 4.19E+17 n/cm 2 in position E-2, 3.70E+07 n/cm 2 s and 6.74E+17 n/cm 2 in position J-8, 2.19E+12 n/cm 2 s and 3.99E+22 n/cm 2 in position. Measurement results in the position B-3 are 2.12E+12 n/cm 2 s and 3.86E+22 n/cm 2 in position G-10 respectively. Other result are fluxes and fluence in beryllium block, those are 5,02E+07 n/cm 2 s and 9,15E+17 n/cm 2 in position (5-6), and 2,32E+07 n/cm 2 s and 4,23E+17 n/cm 2 in position (C-D). Deformation (L/L) results for beryllium element are 1,12E-08 in position E-2, 1,84E-08 in position J-8, 1,60E-03 in position B-3, and 1,55E-03 in position G-10. In beryllium block deformation results are 2,52E-08 in position (5-6) and 1,13E-08 in position (C-D). Those results are shown unseen deformation in beryllium element and beryllium block and demonstrably by visual observation in reactor hot cell. (author)

Beryllium processing technology review for applications in plasma-facing components

Energy Technology Data Exchange (ETDEWEB)

Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

1993-07-01

Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included.

Beryllium processing technology review for applications in plasma-facing components

International Nuclear Information System (INIS)

Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

1993-07-01

Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included

NUCLEAR REACTOR

Science.gov (United States)

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

Proton irradiation effects on beryllium – A macroscopic assessment

Energy Technology Data Exchange (ETDEWEB)

Simos, Nikolaos, E-mail: simos@bnl.gov [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Elbakhshwan, Mohamed [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Zhong, Zhong [Photon Sciences, NSLS II, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Camino, Fernando [Center for Functional Nanomaterials, Brookhaven National Laboratory, Upton, NY, 11973 (United States)

2016-10-15

Beryllium, due to its excellent neutron multiplication and moderation properties, in conjunction with its good thermal properties, is under consideration for use as plasma facing material in fusion reactors and as a very effective neutron reflector in fission reactors. While it is characterized by unique combination of structural, chemical, atomic number, and neutron absorption cross section it suffers, however, from irradiation generated transmutation gases such as helium and tritium which exhibit low solubility leading to supersaturation of the Be matrix and tend to precipitate into bubbles that coalesce and induce swelling and embrittlement thus degrading the metal and limiting its lifetime. Utilization of beryllium as a pion production low-Z target in high power proton accelerators has been sought both for its low Z and good thermal properties in an effort to mitigate thermos-mechanical shock that is expected to be induced under the multi-MW power demand. To assess irradiation-induced changes in the thermal and mechanical properties of Beryllium, a study focusing on proton irradiation damage effects has been undertaken using 200 MeV protons from the Brookhaven National Laboratory Linac and followed by a multi-faceted post-irradiation analysis that included the thermal and volumetric stability of irradiated beryllium, the stress-strain behavior and its ductility loss as a function of proton fluence and the effects of proton irradiation on the microstructure using synchrotron X-ray diffraction. The mimicking of high temperature irradiation of Beryllium via high temperature annealing schemes has been conducted as part of the post-irradiation study. This paper focuses on the thermal stability and mechanical property changes of the proton irradiated beryllium and presents results of the macroscopic property changes of Beryllium deduced from thermal and mechanical tests.

Investigation of reactivity variations of the Isfahan MNSR reactor due to variations in the thickness of the core top beryllium layer using WIMSD and MCNP codes

Directory of Open Access Journals (Sweden)

A Shirani

2010-12-01

Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core and also after 14 years of operation (total operation time of the reactor is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

Mechanisms of hydrogen retention in metallic beryllium and beryllium oxide and properties of ion-induced beryllium nitride; Rueckhaltemechanismen fuer Wasserstoff in metallischem Beryllium und Berylliumoxid sowie Eigenschaften von ioneninduziertem Berylliumnitrid

Energy Technology Data Exchange (ETDEWEB)

Oberkofler, Martin

2011-09-22

In the framework of this thesis laboratory experiments on atomically clean beryllium surfaces were performed. They aim at a basic understanding of the mechanisms occurring upon interaction of a fusion plasma with a beryllium first wall. The retention and the temperature dependent release of implanted deuterium ions are investigated. An atomistic description is developed through simulations and through the comparison with calculations based on density functional theory. The results of these investigations are compared to the behaviour of hydrogen upon implantation into thermally grown beryllium oxide layers. Furthermore, beryllium nitride is produced by implantation of nitrogen into metallic beryllium and its properties are investigated. The results are interpreted with regard to the use of beryllium in a fusion reactor. (orig.)

THE IDAHO NATIONAL LABORATORY BERYLLIUM TECHNOLOGY UPDATE

International Nuclear Information System (INIS)

Glen R. Longhurst

2007-01-01

A Beryllium Technology Update meeting was held at the Idaho National Laboratory on July 18, 2007. Participants came from the U.S., Japan, and Russia. There were two main objectives of this meeting. One was a discussion of current technologies for beryllium in fission reactors, particularly the Advanced Test Reactor and the Japan Materials Test Reactor, and prospects for material availability in the coming years. The second objective of the meeting was a discussion of a project of the International Science and Technology Center regarding treatment of irradiated beryllium for disposal. This paper highlights discussions held during that meeting and major conclusions reached

Status of material development for lifetime expansion of beryllium reflector

Energy Technology Data Exchange (ETDEWEB)

Dorn, C [Materion Brush Beryllium and Composites, California (United States); Tsuchiya, Kunihiko; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Hatano, Y [Univ. of Toyama, Toyama (Japan); Chakrov, P [INP-KNNC, Almaty (Kazakhstan); Kodama, M [Nippon Nuclear Fuel Development Co., Ltd., Oarai, Ibaraki (Japan)

2012-03-15

Beryllium has been used as the reflector element material in the reactor, specifically S-200F structural grade beryllium manufactured by Materion Brush Beryllium and Composites (former, Brush Wellman Inc.). As a part of the reactor upgrade, the Japan Atomic Energy Agency (JAEA) also has carried out the cooperation experiments to extend the operating lifetime of the beryllium reflector elements. It will first be necessary to determine which of the material's physical, mechanical and chemical properties will be the most influential on that choice. The irradiation testing plans to evaluate the various beryllium grades are also briefly considered and prepared. In this paper, material selection, irradiation test plan and PEI development for lifetime expansion of beryllium are described for material testing reactors. (author)

Corrosion of beryllium

International Nuclear Information System (INIS)

Mueller, J.J.; Adolphson, D.R.

1987-01-01

The corrosion behavior of beryllium in aqueous and elevated-temperature oxidizing environments has been extensively studied for early-intended use of beryllium in nuclear reactors and in jet and rocket propulsion systems. Since that time, beryllium has been used as a structural material in les corrosive environments. Its primary applications include gyro systems, mirror and reentry vehicle structures, and aircraft brakes. Only a small amount of information has been published that is directly related to the evaluation of beryllium for service in the less severe or normal atmospheric environments associated with these applications. Despite the lack of published data on the corrosion of beryllium in atmospheric environments, much can be deduced about its corrosion behavior from studies of aqueous corrosion and the experiences of fabricators and users in applying, handling, processing, storing, and shipping beryllium components. The methods of corrosion protection implemented to resist water and high-temperature gaseous environments provide useful information on methods that can be applied to protect beryllium for service in future long-term structural applications

Beryllium R and D for fusion applications

International Nuclear Information System (INIS)

Scaffidi-Argentina, F.; Longhurst, G.R.; Shestakov, V.; Kawamura, H.

2000-01-01

Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal

Beryllium R and D for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Scaffidi-Argentina, F. E-mail: francesco.scaffidi@iket.fzk.de; Longhurst, G.R.; Shestakov, V.; Kawamura, H

2000-11-01

Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal.

Some aspects of beryllium disposal in Kazakhstan

International Nuclear Information System (INIS)

Shestakov, V.; Chikhray, Y.; Shakhvorostov, Yr.

2004-01-01

Historically in Kazakhstan all disposals of used beryllium and beryllium wasted materials were stored and recycled at JSC ''Ulba Metallurgical Plant''. Since Ulba Metallurgical Plant (beside beryllium and tantalum production) is one of the world largest complex producers of fuel for nuclear power plants as well it has possibilities, technologies and experience in processing toxic and radioactive wastes related with those productions. At present time only one operating Kazakhstan research reactors (EWG1M in Kurchatov) contains beryllium made core. The results of current examination of that core allow using it without replacement long time yet (at least for next five-ten years). Nevertheless the problem how to utilize such irradiated beryllium becomes actual issue for Kazakhstan even today. Since Kazakhstan is the member of ITER/DEMO Reactors Projects and is permanently considered as possible provider of huge amount of beryllium for those reactors so that is the reason for starting studies of possibilities of large scale processing/recycling of such disposed irradiated beryllium. It is clear that the Ulba Metallurgical Plant is considered as the best site for it in Kazakhstan. The draft plan how to organize experimental studies of irradiated beryllium disposals in Kazakhstan involving National Nuclear Center, National University (Almaty), JSC ''Ulba Metallurgical Plant'' (Ust-Kamenogorsk) would be presented in this paper as well as proposals to arrange international collaboration in that field through ISTC (International Science Technology Center, Moscow). (author)

Beryllium technology workshop, Clearwater Beach, Florida, November 20, 1991

International Nuclear Information System (INIS)

Longhurst, G.R.

1991-12-01

This report discusses the following topics: beryllium in the ITER blanket; mechanical testing of irradiated beryllium; tritium release measurements on irradiated beryllium; beryllium needs for plasma-facing components; thermal conductivity of plasma sprayed beryllium; beryllium research at the INEL; Japanese beryllium research activities for in-pile mockup tests on ITER; a study of beryllium bonding of copper alloy; new production technologies; thermophysical properties of a new ingot metallurgy beryllium product line; implications of beryllium:steam interactions in fusion reactors; and a test program for irradiation embrittlement of beryllium at JET

Beryllium technology workshop, Clearwater Beach, Florida, November 20, 1991

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R.

1991-12-01

This report discusses the following topics: beryllium in the ITER blanket; mechanical testing of irradiated beryllium; tritium release measurements on irradiated beryllium; beryllium needs for plasma-facing components; thermal conductivity of plasma sprayed beryllium; beryllium research at the INEL; Japanese beryllium research activities for in-pile mockup tests on ITER; a study of beryllium bonding of copper alloy; new production technologies; thermophysical properties of a new ingot metallurgy beryllium product line; implications of beryllium:steam interactions in fusion reactors; and a test program for irradiation embrittlement of beryllium at JET.

The Dragon reactor experiment

International Nuclear Information System (INIS)

Anon.

1976-01-01

The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af

Experiments on tritium behavior in beryllium, (1)

International Nuclear Information System (INIS)

Kawamura, Hiroshi; Ishizuka, Etsuo; Matsumoto, Mikio; Inada, Seiji; Sezaki, Katsuji; Saito, Minoru; Kato, Mineo.

1989-06-01

In JMTR, it was observed that the tritium concentration of the primary coolant increases with the reactor operation at 50 MW. As one of the tritium generation sources, we paid attention to a neutron reflector made of beryllium because the tritium generation rate in the beryllium is bigger than other components in the reactor core. On the other hand, the irradiation test of blanket materials (i.e. tritium breeding materials and neutron multipling materials) are planned for development of the fusion reactor in JMTR and the beryllium will be also irradiated as a neutron multiplier with tritium breeding materials. Therefore, as the irradiated specimens, we used a hot-pressed beryllium disk fabricated by the same method as the neutron reflector or the neutron multiplier and conducted the irradiation tests in JMTR. The purpose of these tests are to clarify the tritium behavior in the hot-pressed beryllium. In this paper, from a viewpoint of the fabrication of capsules for neutron irradiation, the specifications of the irradiated specimens and capsules are summarized. Additionally, the results on the puncture test of the container of the irradiation specimens are described. (author)

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [SCK CEN (Belgium); Kalcheva, S. [SCK CEN (Belgium); Sikik, E. [SCK CEN (Belgium); Koonen, E. [SCK CEN (Belgium)

2015-12-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water (Figure 1). The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm2 to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident.

The status of beryllium technology for fusion

Energy Technology Data Exchange (ETDEWEB)

Scaffidi-Argentina, F.; Longhurst, G.R. E-mail: gx1@inel.gov; Shestakov, V.; Kawamura, H

2000-12-01

Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented.

The status of beryllium technology for fusion

International Nuclear Information System (INIS)

Scaffidi-Argentina, F.; Longhurst, G.R.; Shestakov, V.; Kawamura, H.

2000-01-01

Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented

Experiments on tritium behavior in beryllium, (2)

International Nuclear Information System (INIS)

Ishitsuka, Etsuo; Kawamura, Hiroshi; Nakata, Hirokatsu; Sugai, Hiroyuki; Tanase, Masakazu.

1990-02-01

Beryllium has been used as the neutron reflector of material testing reactor and as the neutron multiplier for the fusion reactor lately. To study the tritium behavior in beryllium, we conducted the experiments, i.e., tritium release by recoil or diffusion by using the hot-pressed beryllium which had been produced both tritium and helium by neutron irradiation. From our experiments, we found that (1) amount of tritium production per one cycle irradiation (lasting 22 days) of JMTR is 10 mCi/g, (2) amount of tritium per surface area of hot-pressed beryllium released by recoil is 4 μCi/cm 2 , (3) diffusion coefficient of tritium in a temperature range of 800 ∼1180degC can be expressed with the following equation; D = 8.7 x 10 4 exp(-2.9x10 5 /R/T) cm 2 /s. (author)

Research of beryllium safety issues

International Nuclear Information System (INIS)

Longhurst, G.R.; Anderl, R.A.; Dolan, T.J.; Hankins, M.R.; Pawelko, R.J.

1993-01-01

Beryllium has been identified as a leading contender for the plasma-facing material in ITER. Its use has some obvious advantages, but there are also a number of safety concerns associated with it. The Idaho National Engineering Laboratory (INEL) has undertaken a number of studies to help resolve some of these issues. One issue is the response of beryllium to neutron irradiation. We have tested samples irradiated in the Advanced Test Reactor (ATR) and are currently preparing to make measurements of the change in mechanical properties of beryllium samples irradiated at elevated temperatures in the Fast Flux Test Facility (FFTF) and the Experimental Breeder Reactor II (EBR-II) at the INEL. Mechanical tests will be conducted at the irradiation temperatures of 375-550 C. Other experiments address permeation and retention of implanted tritium in plasma-sprayed beryllium. In one test the porosity of the material allowed 0.12% of implanted ions and 0.17% of atoms from background gas pressure to pass through the foil with essentially no delay. For comparison, similar tests on fully dense hot-rolled, vacuum melted or sintered powder foils of high purity beryllium showed only 0.001% of implanting ions to pass through the foil, and then only after a delay of several hours. None of the molecular gas appeared to permeate these latter targets. An implication is that plasma-sprayed beryllium may substantially enhance recycling of tritium to the plasma provided it is affixed to a relatively impermeable substrate. (orig.)

Research of flaw assessment methods for beryllium reflector elements

International Nuclear Information System (INIS)

Shibata, Akira; Ito, Masayasu; Takemoto, Noriyuki; Tanimoto, Masataka; Tsuchiya, Kunihiko; Nakatsuka, Masafumi; Ohara, Hiroshi; Kodama, Mitsuhiro

2012-02-01

Reflector elements made from metal beryllium is widely used as neutron reflectors to increase neutron flux in test reactors. When beryllium reflector elements are irradiated by neutron, bending of reflector elements caused by swelling occurs, and beryllium reflector elements must be replaced in several years. In this report, literature search and investigation for non-destructive inspection of Beryllium and experiments for Preliminary inspection to establish post irradiation examination method for research of characteristics of metal beryllium under neutron irradiation were reported. (author)

Safety handling of beryllium for fusion technology R and D

International Nuclear Information System (INIS)

Yoshida, Hiroshi; Okamoto, Makoto; Terai, Takayuki; Odawara, Osamu; Ashibe, Kusuo; Ohara, Atsushi.

1992-07-01

Feasibility of beryllium use as a blanket neutron multiplier, first wall and plasma facing material has been studied for the D-T burning experiment reactors such as ITER. Various experimental work of beryllium and its compounds will be performed under the conditions of high temperature and high energy particle exposure simulating fusion reactor conditions. Beryllium is known as a hazardous substance and its handling has been carefully controlled by various health and safe guidances and/or regulations in many countries. Japanese regulations for hazardous substance provide various guidelines on beryllium for the protection of industrial workers and environment. This report was prepared for the safe handling of beryllium in a laboratory scale experiments for fusion technology R and D such as blanket development. Major items in this report are; (1) Brief review of guidances and regulations in USA, UK and Japan. (2) Safe handling and administration manuals at beryllium facilities in INEL, LANL and JET. (3) Conceptual design study of beryllium handling facility for small to mid-scale blanket R and D. (4) Data on beryllium toxicity, example of clinical diagnosis of beryllium disease, and environmental occurence of beryllium. (5) Personnel protection tools of Japanese Industrial Standard for hazardous substance. (author) 61 refs

HEINBE; the calculation program for helium production in beryllium under neutron irradiation

International Nuclear Information System (INIS)

Shimakawa, Satoshi; Ishitsuka, Etsuo; Sato, Minoru

1992-11-01

HEINBE is a program on personal computer for calculating helium production in beryllium under neutron irradiation. The program can also calculate the tritium production in beryllium. Considering many nuclear reactions and their multi-step reactions, helium and tritium productions in beryllium materials irradiated at fusion reactor or fission reactor may be calculated with high accuracy. The calculation method, user's manual, calculated examples and comparison with experimental data were described. This report also describes a neutronics simulation method to generate additional data on swelling of beryllium, 3,000-15,000 appm helium range, for end-of-life of the proposed design for fusion blanket of the ITER. The calculation results indicate that helium production for beryllium sample doped lithium by 50 days irradiation in the fission reactor, such as the JMTR, could be achieved to 2,000-8,000 appm. (author)

Compact power reactor

International Nuclear Information System (INIS)

Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

Thermogravimetric analysis of the beryllium/steam reaction

Energy Technology Data Exchange (ETDEWEB)

Druyts, Frank E-mail: fdruyts@sckcen.be; Iseghem, Pierre van

2000-11-01

In view of the safety assessment of new fusion reactor designs, kinetic data are needed on the beryllium/steam reaction. Therefore, thermogravimetric analysis was used to determine the reactivity of beryllium in steam as a function of temperature, irradiation history and porosity of the samples. To this purpose, reference unirradiated S-200 VHP beryllium samples were compared with specimens irradiated in the BR2 reactor up to fast neutron fluences (E>1 MeV) of respectively 1.6x10{sup 21} n cm{sup -2} (resulting in a helium content of 300 appm He and a theoretical density of 99.9%) and 4x10{sup 22} n cm{sup -2} (21000 appm He, 97.2% theoretical density). Kinetics were parabolic for all tested beryllium types at 600 deg. C. At 700 deg. C, kinetics were parabolic for the unirradiated and irradiated 99.9% dense beryllium, and accelerating/linear for the irradiated 97.2% material. At 800 deg. C, all samples showed accelerating/linear behaviour. There was no influence of porosity on the reaction rate of beryllium in steam within the limited investigated density range, except at 700 deg. C, where the measured reaction rate for the irradiated 97.2% dense samples is an order of magnitude higher than for the irradiated 99.9% dense specimens.

Reaction-diffusion modeling of hydrogen in beryllium

Energy Technology Data Exchange (ETDEWEB)

Wensing, Mirko; Matveev, Dmitry; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik (Germany)

2016-07-01

Beryllium will be used as first-wall material for the future fusion reactor ITER as well as in the breeding blanket of DEMO. In both cases it is important to understand the mechanisms of hydrogen retention in beryllium. In earlier experiments with beryllium low-energy binding states of hydrogen were observed by thermal desorption spectroscopy (TDS) which are not yet well understood. Two candidates for these states are considered: beryllium-hydride phases within the bulk and surface effects. The retention of deuterium in beryllium is studied by a reaction rate approach using a coupled reaction diffusion system (CRDS)-model relying on ab initio data from density functional theory calculations (DFT). In this contribution we try to assess the influence of surface recombination.

The beryllium production at Ulba metallurgical plant (Ust-Kamenogrsk, Kazakhstan)

Energy Technology Data Exchange (ETDEWEB)

Dvinskykh, E.M.; Savchuk, V.V.; Tuzov, Y.V. [Ulba Metallurgical Plant (Zavod), Ust-Kamenogorsk, Abay prospect 102 (Kazakhstan)

1998-01-01

The Report includes data on beryllium production of Ulba metallurgical plant, located in Ust-Kamenogorsk (Kazakhstan). Beryllium production is showed to have extended technological opportunities in manufacturing semi-products (beryllium ingots, master alloys, metallic beryllium powders, beryllium oxide) and in production of structural beryllium and its parts. Ulba metallurgical plant owns a unique technology of beryllium vacuum distillation, which allows to produce reactor grades of beryllium with a low content of metallic impurities. At present Ulba plant does not depend on raw materials suppliers. The quantity of stored raw materials and semi-products will allow to provide a 25-years work of beryllium production at a full capacity. The plant has a satisfactory experience in solving ecological problems, which could be useful in ITER program. (author)

New facility for post irradiation examination of neutron irradiated beryllium

International Nuclear Information System (INIS)

Ishitsuka, Etsuo; Kawamura, Hiroshi

1995-01-01

Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

Optimization of beryllium for fusion blanket applications

International Nuclear Information System (INIS)

Billone, M.C.

1993-01-01

The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

Reducing the cost of S-65C grade beryllium for ITER first wall applications

International Nuclear Information System (INIS)

Kaczynski, D.; Sato, K.; Savchuk, V.V.; Shestakov, V.P.

2004-01-01

Beryllium is the current material of choice for plasma-facing components in ITER. The present design is for 10 mm thick beryllium tiles bonded to an actively cooled copper substrate. Brush Wellman grade S65C beryllium is preferred grade off beryllium for these tiles. S65C has the best resistance to low-cycle thermal fatigue than any other beryllium grad in the world. S65C grade beryllium has been successfully deployed in fusion reactors for more than two decades, most recently in the JET reactor. This paper will detail a supply chain to produce the most cost-effective S65C plasma facing components for ITER. This paper will also propose some future work too demonstrate the best technology for bonding beryllium to copper. (author)

Hydrogen transport behavior of beryllium

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; Hankins, M.R.; Longhurst, G.R.; Pawelko, R.J. (Idaho National Engineering Lab., EG and G Idaho, Inc., Idaho Falls, ID (United States)); Macaulay-Newcombe, R.G. (Dept. of Engineering Physics, Univ. Hamilton, ON (Canada))

1992-12-01

Beryllium is being evaluated for use as a plasma-facing material in the International Thermonuclear Experimental Reactor (ITER). One concern in the evaluation is the retention and permeation of tritium implanted into the plasma-facing surface. We performed laboratory-scale studies to investigate mechanisms that influence hydrogen transport and retention in beryllium foil specimens of rolled powder metallurgy product and rolled ingot cast beryllium. Specimen characterization was accomplished using scanning electron microscopy. Auger electron spectroscopy, and Rutherford backscattering spectrometry (RBS) techniques. Hydrogen transport was investigated using ion-beam permeation experiments and nuclear reaction analysis (NRA). Results indicate that trapping plays a significant role in permeation, re-emission, and retention, and that surface processes at both upstream and downstream surfaces are also important. (orig.).

High dose neutron irradiation damage in beryllium as blanket material

Energy Technology Data Exchange (ETDEWEB)

Chakin, V.P. E-mail: fae@niiar.ru; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B. E-mail: vniinm.400@g23.relkom.ru

2001-11-01

The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10{sup 22} and 8.0x10{sup 22} cm{sup -2} (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10{sup 22} cm{sup -2} (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10{sup 22} cm{sup -2} (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

High dose neutron irradiation damage in beryllium as blanket material

International Nuclear Information System (INIS)

Chakin, V.P.; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B.

2001-01-01

The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10 22 and 8.0x10 22 cm -2 (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10 22 cm -2 (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10 22 cm -2 (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket

Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

2012-07-01

The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

Interaction of hydrogen and its isotopes with irradiated beryllium

International Nuclear Information System (INIS)

Tazhibaeva, I.L.; Shestakov, V.P.; Klepikov, A.Kh.; Pomanenko, O.G.; Chikhraj, E.V.; Kenzhin, E.A.; Zverev, V.V.; Kolbanenkov, A.N.

2000-01-01

In the article the results of experiments on hydrogen and its isotopes accumulation and gas-release from irradiated beryllium are presented. The irradiation was conducted at different media and temperatures in the RA and IVG.1M reactors. The measurements were carried out by thermal desorption method. Hydrogen release from beryllium samples saturated at different conditions were calculated. Dependence of hydrogen confinement character in beryllium from grain orientation in the sample, temperature and irradiation rate was revealed

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-09-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm² to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

Thermal conductivity of beryllium under low temperature high dose neutron irradiation

International Nuclear Information System (INIS)

Chakin, V.P.; Latypov, R.N.; Suslov, D.N.; Kupriyanov, I.B.

2004-01-01

Thermal conductivity of compact beryllium of several Russian grades such as TE-400, TE-56, TE-30, TIP and DIP differing in the production technology, grain size and impurity content has been investigated. The thermal diffusivity of beryllium was measured on the disks in the initial and irradiated conditions using the pulse method in the range from room temperature to 200degC. The thermal conductivity was calculated using the table values for the beryllium thermal capacity. The specimens and beryllium neutron source fragments were irradiation in the SM reactor at 70degC and 200degC to a neutron fluence of (0.5-11.4)·10 22 cm -2 (E>0.1 MeV) and in the BOR-60 reactor at 400degC to 16·10 22 cm -2 (E>0.1MeV), respectively. The low-temperature irradiation leads to the drop decrease of the beryllium thermal conductivity and the effect depends on the irradiation parameters. The paper analyses the effect of irradiation parameters (temperature, neutron fluence), measurement temperature and structural factors on beryllium conductivity. The experiments have revealed that the short time post-irradiation annealing at high temperature results in partial reduction of the thermal conductivity of irradiated beryllium. (author)

MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

Science.gov (United States)

Balent, R.

1963-03-12

This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

Overview moderator material for nuclear reactor components

International Nuclear Information System (INIS)

Mairing Manutu Pongtuluran; Hendra Prihatnadi

2009-01-01

In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

TEM study of impurity segregations in beryllium pebbles

Energy Technology Data Exchange (ETDEWEB)

Klimenkov, M., E-mail: michael.klimenkov@kit.edu [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Chakin, V.; Moeslang, A. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R. [Institute for Applied Materials – Materials and Biomechanics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

2014-12-15

Beryllium is planned to be used as a neutron multiplier in the Helium-cooled Pebble Bed European concept of a breeding blanket of demonstration power reactor DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron-irradiated at temperatures typical of fusion blankets. Beryllium pebbles 1 mm in diameter produced by the rotating electrode method were subjected to a TEM study before and after irradiation at High Flux Reactor, Petten, Netherlands at 861 K. The grain size varied in a wide range from sub-micron size up to several tens of micrometers, which indicated formation bimodal grain size distribution. Based on the application of combined electron energy loss spectroscopy and energy dispersive X-ray spectroscopy methods, we suggest that impurity precipitates play an important role in controlling the mechanical properties of beryllium. The impurity elements were present in beryllium at a sub-percent concentration form beryllide particles of a complex (Fe/Al/Mn/Cr)B composition. These particles are often ordered along dislocations lines, forming several micron-long chains. It can be suggested that fracture surfaces often extended along these chains in irradiated material.

Behavior of beryllium pebbles under irradiation

Energy Technology Data Exchange (ETDEWEB)

Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

1998-01-01

Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

Particle bed reactor nuclear rocket concept

International Nuclear Information System (INIS)

Ludewig, H.

1991-01-01

The particle bed reactor nuclear rocket concept consists of fuel particles (in this case (U,Zr)C with an outer coat of zirconium carbide). These particles are packed in an annular bed surrounded by two frits (porous tubes) forming a fuel element; the outer one being a cold frit, the inner one being a hot frit. The fuel element are cooled by hydrogen passing in through the moderator. These elements are assembled in a reactor assembly in a hexagonal pattern. The reactor can be either reflected or not, depending on the design, and either 19 or 37 elements, are used. Propellant enters in the top, passes through the moderator fuel element and out through the nozzle. Beryllium used for the moderator in this particular design to withstand the high radiation exposure implied by the long run times

Moderator heat recovery of CANDU reactors

International Nuclear Information System (INIS)

Fath, H.E.S.; Ahmed, S.T.

1986-01-01

A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

Measuring device for bending of beryllium reflector

International Nuclear Information System (INIS)

Nishida, Seiri; Sakamoto, Naoki.

1994-01-01

The device of the present invention can measure bending of a beryllium reflector formed in a reactor core of a nuclear reactor by a relatively easy operation. Namely, a sensor portion comprises a long-support that can be inserted to a fuel element-insertion hole disposed in the reactor and a plurality of distance sensors disposed in a longitudinal direction of the support. A supersonic wave sensor which is advantageous in the heat resistance, the size and the accuracy and can conduct measurement in water relatively easily is used as the distance sensors. However, other sensors, instead of the sensor described above, may also be used. The plurality of distance sensors detect the bending amount of the beryllium reflector in the longitudinal direction by such an easy operation of inserting such a sensor portion to the fuel element-insertion hole upon exchange of fuel elements. (I.S.)

Mechanical performance of irradiated beryllium pebbles

Energy Technology Data Exchange (ETDEWEB)

Scaffidi-Argentina, F.; Dalle-Donne, M.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

1998-01-01

For the Helium Cooled Pebble Bed (HCPB) Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier consists of a mixed bed of about 2 and 0.1-0.2 mm diameter beryllium pebbles. Beryllium has no structural function in the blanket, however microstructural and mechanical properties are important, as they might influence the material behavior under neutron irradiation. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating it. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation of essential material properties and trying to elucidate the processes controlling the property changes. The microstructure, the porosity distribution, the impurity content, the behavior under compression loads and the compatibility of the beryllium pebbles with lithium orthosilicate (Li{sub 4}SiO{sub 4}) during the in-pile irradiation are presented and critically discussed. Qualitative information on ductility and creep obtained by hardness-type measurements are also supplied. (author)

Study on neutron irradiation behavior of beryllium as neutron multiplier

Energy Technology Data Exchange (ETDEWEB)

Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-03-01

More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

Beryllium data base for in-pile mockup test on blanket of fusion reactor, (1)

Energy Technology Data Exchange (ETDEWEB)

Kawamura, Hiroshi; Ishitsuka, Etsuo (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Sakamoto, Naoki; Kato, Masakazu; Takatsu, Hideyuki.

1992-11-01

Beryllium has been used in the fusion blanket designs with ceramic breeder as a neutron multiplier to increase the net tritium breeding ratio (TBR). The properties of beryllium, that is physical properties, chemical properties, thermal properties, mechanical properties, nuclear properties, radiation effects, etc. are necessary for the fusion blanket design. However, the properties of beryllium have not been arranged for the fusion blanket design. Therefore, it is indispensable to check and examine the material data of beryllium reported previously. This paper is the first one of the series of papers on beryllium data base, which summarizes the reported material data of beryllium. (author).

Reactivity effects due to beryllium poisoning of BR2

International Nuclear Information System (INIS)

Kalcheva, S.; Ponsard, B.; Koonen, E.

2004-01-01

This paper illustrates the impact of the poisoning of the beryllium reflector on reactivity variations of the Belgian MTR BR2 in SCK.CEN. Detailed calculations by MCNP-4C of reactivity effects caused by strong neutron absorbers 3 He and 6 Li during reactor operation history are presented. The importance of beryllium poisoning for the accuracy of reactivity predictions is discussed. (authors)

Beryllium

Science.gov (United States)

Foley, Nora K.; Jaskula, Brian W.; Piatak, Nadine M.; Schulte, Ruth F.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

2017-12-19

Beryllium is a mineral commodity that is used in a variety of industries to make products that are essential for the smooth functioning of a modern society. Two minerals, bertrandite (which is supplied domestically) and beryl (which is currently supplied solely by imports), are necessary to ensure a stable supply of high-purity beryllium metal, alloys, and metal-matrix composites and beryllium oxide ceramics. Although bertrandite is the source mineral for more than 90 percent of the beryllium produced globally, industrial beryl is critical for the production of the very high purity beryllium metal needed for some strategic applications. The current sole domestic source of beryllium is bertrandite ore from the Spor Mountain deposit in Utah; beryl is imported mainly from Brazil, China, Madagascar, Mozambique, and Portugal. High-purity beryllium metal is classified as a strategic and critical material by the Strategic Materials Protection Board of the U.S. Department of Defense because it is used in products that are vital to national security. Beryllium is maintained in the U.S. stockpile of strategic materials in the form of hot-pressed beryllium metal powder.Because of its unique chemical properties, beryllium is indispensable for many important industrial products used in the aerospace, computer, defense, medical, nuclear, and telecommunications industries. For example, high-performance alloys of beryllium are used in many specialized, high-technology electronics applications, as they are energy efficient and can be used to fabricate miniaturized components. Beryllium-copper alloys are used as contacts and connectors, switches, relays, and shielding for everything from cell phones to thermostats, and beryllium-nickel alloys excel in producing wear-resistant and shape-retaining high-temperature springs. Beryllium metal composites, which combine the fabrication ability of aluminum with the thermal conductivity and highly elastic modulus of beryllium, are ideal for

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

Verpoortem, J. R.; Van Dyck, S.

2014-01-01

At the Belgian nuclear research centre (SCK.CEN) several test reactors are operated. Among these, Belgian Reactor 2 (BR2) is the largest Material Test Reactor (MTR). This water-cooled, beryllium moderated reactor with a maximum thermal power of 100 MW became operational in 1962. Except for two major refurbishment campaigns of one year each, this reactor has been operated continuously over the past 50 years, with a frequency of 5-12 cycles per year. At present, BR2 is used for different research activities, the production of medical isotopes, the production of n-doped silicon and various training and education activities. (Author)

Neutron irradiation behavior of ITER candidate beryllium grades

Energy Technology Data Exchange (ETDEWEB)

Kupriyanov, I.B.; Gorokhov, V.A.; Nikolaev, G.N. [A.A.Bochvar All-Russia Scientific Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Melder, R.R.; Ostrovsky, Z.E.

1998-01-01

Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 {mu}m, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) {center_dot} 10{sup 21} cm{sup -2} (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T{sub 1} = 130-180degC and T{sub 2} = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 {center_dot} 10{sup 21} cm{sup -2}. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

Mechanical properties of irradiated beryllium

International Nuclear Information System (INIS)

Beeston, J.M.; Longhurst, G.R.; Wallace, R.S.

1992-01-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10 25 n/m 2 (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.)

Mechanical properties of irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Beeston, J.M.; Longhurst, G.R.; Wallace, R.S. (EG and G Idaho, Inc., Idaho Falls, ID (United States). Idaho National Engineering Lab.); Abeln, S.P. (EG and G Rocky Flats, Inc., Golden, CO (United States))

1992-10-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10[sup 25] n/m[sup 2] (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.).

Mechanical properties of irradiated beryllium

Science.gov (United States)

Beeston, J. M.; Longhurst, G. R.; Wallace, R. S.; Abeln, S. P.

1992-10-01

Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 × 10 25 n/m 2 ( E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.

Workshop on beryllium for fusion applications. Proceedings. IEA Implementing Agreement for a Programme of Research and Development on Fusion Materials

International Nuclear Information System (INIS)

Dalle Donne, M.

1993-12-01

As shown by recent developments beryllium has become one of the most important materials in the development of fusion reactors. It is practically the only neutron multiplier available for blankets with ceramic breeder materials and can be used with liquid metal breeders as well. It is one of the most likely materials to be used on the surface of the first walls and of the divertor. The neutron irradiation behavior of beryllium in a fusion reactor is not well know. Beryllium was extensively irradiated about 25-40 years ago and has been used since then in material testing reactors as reflector. In the meantime, however, beryllium has been improved quite considerably. Today it is possible to obtain commercially beryllium which is much more isotropic and contains smaller ammounts of oxide. There are already indications that these new kinds of beryllium behave better under irradiation. (orig.)

The structure, properties and performance of plasma-sprayed beryllium for fusion applications

International Nuclear Information System (INIS)

Castro, R.G.; Stanek, P.W.; Elliott, K.E.

1995-01-01

Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasma-spray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H 2 gas additions to improve the melting of the beryllium powder and transferred-arc cleaning to improve the bonding between the beryllium coatings and the underlying surface. Information will also be presented on thermal fatigue tests which were done on beryllium coated ISX-B beryllium limiter tiles using 10 sec cycle times with 60 sec cooldowns and an International Thermonuclear Experimental Reactor (ITER) relevant divertor heat flux slightly in excess of 5 MW/m 2

In-pile thermocycling testing and post-test analysis of beryllium divertor mockups

Energy Technology Data Exchange (ETDEWEB)

Giniatulin, R.; Mazul, I. [Efremov Inst., St. Petersburg (Russian Federation); Melder, R.; Pokrovsky, A.; Sandakov, V.; Shiuchkin, A.

1998-01-01

The main damaging factors which impact the ITER divertor components are neutron irradiation, cyclic surface heat loads and hydrogen environment. One of the important questions in divertor mockups development is the reliability of beryllium/copper joints and the beryllium resistance under neutron irradiation and thermal cycling. This work presents the experiment, where neutron irradiation and thermocyclic heat loads were applied simultaneously for two beryllium/copper divertor mockups in a nuclear reactor channel to simulate divertor operational conditions. Two mockups with different beryllium grades were mounted facing each other with the tantalum heater placed between them. This device was installed in the active zone of the nuclear reactor SM-2 (Dimitrovgrad, Russia) and the tantalum block was heated by neutron irradiation up to a high temperature. The main part of the heat flux from the tantalum surface was transported to the beryllium surface through hydrogen, as a result the heat flux loaded two mockups simultaneously. The mockups were cooled by reactor water. The device was lowered to the active zone so as to obtain the heating regime and to provide cooling lifted. This experiment was performed under the following conditions: tantalum heater temperature - 1950degC; hydrogen environment -1000 Pa; surface heat flux density -3.2 MW/m{sup 2}; number of thermal cycles (lowering and lifting) -101; load time in each cycle - 200-5000 s; dwell time (no heat flux, no neutrons) - 300-2000 s; cooling water parameters: v - 1 m/s, Tin - 50degC, Pin - 5 MPa; neutron fluence -2.5 x 10{sup 20} cm{sup -2} ({approx}8 years of ITER divertor operation from the start up). The metallographic analysis was performed after experiment to investigate the beryllium and beryllium/copper joint structures, the results are presented in the paper. (author)

The status of beryllium research for fusion in the United States

International Nuclear Information System (INIS)

Longhurst, G.R.; Snead, L.L.; Abou-Sena, A.A.

2004-01-01

Use of beryllium in fusion reactor has been considered for neutron multiplication in breeding blankets an as an oxygen getter for plasma - facing surface. Previous beryllium research for fusion in the United States included issues of interest to fission (swelling an changes in mechanical and thermal properties) as well as interactions with plasmas and hydrogen isotopes and methods of fabrication. When the United States formally withdrew its participation in the International Experimental Reactor (ITER) program, much of this effort was terminated. The focus in the U.S. has been mainly on toxic effects of beryllium and on industrial hygiene and health-related issues. Work continued at the INEEL (Idaho National Engineering and Environmental Laboratory) and elsewhere on beryllium-containing molten salts. This activity is part of the JUPITER II Agreement. Plasma spray of ITER first wall samples at Los Alamos National Laboratory has been performed under the European Fusion Development Agreement. Effects of irradiation on beryllium structure are being studied at Oak Ridge National Laboratory. Numerical and phenomenological models are being developed and applied at the University of California Los Angels to investigate thermo-mechanical characteristics of beryllium pebble beds, similar to research being carried out at Forschungszentrum Karlsruhe and elsewhere. Additional work, not funded by the fusion program, has dealt with issues of disposal, and recycling. (author)

The Status of Beryllium Research for Fusion in the United States

International Nuclear Information System (INIS)

Glen R. Longhurst

2003-01-01

Use of beryllium in fusion reactors has been considered for neutron multiplication in breeding blankets and as an oxygen getter for plasma-facing surfaces. Previous beryllium research for fusion in the United States included issues of interest to fission (swelling and changes in mechanical and thermal properties) as well as interactions with plasmas and hydrogen isotopes and methods of fabrication. When the United States formally withdrew its participation in the International Thermonuclear Experimental Reactor (ITER) program, much of this effort was terminated. The focus in the U.S. has been mainly on toxic effects of beryllium and on industrial hygiene and health-related issues. Work continued at the INEEL and elsewhere on beryllium-containing molten salts. This activity is part of the JUPITER II Agreement. Plasma spray of ITER first wall samples at Los Alamos National Laboratory has been performed under the European Fusion Development Agreement. Effects of irradiation on beryllium structure are being studied at Oak Ridge National Laboratory. Numerical and phenomenological models are being developed and applied to better understand important processes and to assist with design. Presently, studies are underway at the University of California Los Angeles to investigate thermo-mechanical characteristics of beryllium pebble beds, similar to research being carried out at Forschungszentrum Karlsruhe and elsewhere. Additional work, not funded by the fusion program, has dealt with issues of disposal, and recycling

Recommended design correlations for S-65 beryllium

International Nuclear Information System (INIS)

Billone, M.C.

1995-01-01

The properties of tritium and helium behavior in irradiated beryllium are reviewed, along with the thermal-mechanical properties needed for ITER design analysis. Correlations are developed to describe the performance of beryllium in a fusion reactor environment. While this paper focuses on the use of beryllium as a plasma-facing component (PFC) material, the correlations presented here can also be used to describe the performance of beryllium as a neutron multiplier for a tritium breeding blanket. The performance properties for beryllium are subdivided into two categories: properties which do not change with irradiation damage to the bulk of the material; and properties which are degraded by neutron irradiation. The irradiation-independent properties described within are: thermal conductivity, specific heat capacity, thermal expansion, and elastic constants. Irradiation-dependent properties include: yield strength, ultimate tensile strength, plastic tangent modulus, uniform and total tensile elongation, thermal and irradiation-induced creep strength, He-induced swelling and tritium retention/release. The approach taken in developing properties correlations is to describe the behavior of dense, pressed S-65 beryllium -- the material chosen for ITER PFC application -- as a function of temperature. As there are essentially no data on the performance of porous and/or irradiated S-65 beryllium, the degradation of properties with as-fabricated porosity and irradiation are determined from the broad data base on S-200F, as well as other types and grades, and applied to S-65 beryllium by scaling factors. The resulting correlations can be used for Be produced by vacuum hot pressing (VHP) and cold-pressing (CP)/sintering(S)/hot-isostatic-pressing (HIP). The performance of plasma-sprayed beryllium is discussed but not quantified

First experience with the new solid methane moderator at the IBR-2 reactor

International Nuclear Information System (INIS)

Beliakov, A.A.; Shabalin, E.P.; Tretyakov, I.T.

2001-01-01

In the 1999 Fall the solid methane moderator (CM) has been installed and tested at full power at the IBR-2 pulsed reactor. Its main features are a beryllium reflector and a light water premoderator. Radiation load on the methane was three times as much as that of IPNS facility, namely, 0.1 W/g. Effects of temperature, operation time, concentration of a hydrogen scavenger, and annealing procedure on both neutron and service performances were studied. Maximum operation time of a newly loaded portion of methane was 4 days. In this time around 30% of methane is transformed into hydrogen, ethane, and high molecular hydrocarbons, and yet no deterioration in cold neutron intensity was detected. Among new knowledge, the most important are two facts observed: two-fold decrease in hydrogen formation rate when methane is poisoned with 2.5% to 5% of ethylene, and low formation rate of solid, inremovable products of radiolysis - (1.5/3)10 -7 g/J, which means that after 10 years of operation the methane chamber will be filled with only 100 g of residue. Gain of factor 20 in cold neutron flux was obtained as compared to the routine grooved light water moderator. Presently, it is the highest among the intense pulsed neutron sources. (author)

Irradiated Beryllium Disposal Workshop, Idaho Falls, ID, May 29-30, 2002

Energy Technology Data Exchange (ETDEWEB)

Longhurst, Glen Reed; Anderson, Gail; Mullen, Carlan K; West, William Howard

2002-07-01

In 2001, while performing routine radioactive decay heat rate calculations for beryllium reflector blocks for the Advanced Test Reactor (ATR), it became evident that there may be sufficient concentrations of transuranic isotopes to require classification of this irradiated beryllium as transuranic waste. Measurements on samples from ATR reflector blocks and further calculations confirmed that for reflector blocks and outer shim control cylinders now in the ATR canal, transuranic activities are about five times the threshold for classification. That situation implies that there is no apparent disposal pathway for this material. The problem is not unique to the ATR. The High Flux Isotope Reactor at Oak Ridge National Laboratory, the Missouri University Research Reactor at Columbia, Missouri and other reactors abroad must also deal with this issue. A workshop was held in Idaho Falls Idaho on May 29-30, 2002 to acquaint stakeholders with these findings and consider a path forward in resolving the issues attendant to disposition of irradiated material. Among the findings from this workshop were (1) there is a real potential for the US to be dependent on foreign sources for metallic beryllium within about a decade; (2) there is a need for a national policy on beryllium utilization and disposition and for a beryllium coordinating committee to be assembled to provide guidance on that policy; (3) it appears it will be difficult to dispose of this material at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico due to issues of Defense classification, facility radioactivity inventory limits, and transportation to WIPP; (4) there is a need for a funded DOE program to seek resolution of these issues including research on processing techniques that may make this waste acceptable in an existing disposal pathway or allow for its recycle.

Beryllium-copper reactivity in an ITER joining environment

International Nuclear Information System (INIS)

Odegard, B.C.; Cadden, C.H.; Yang, N.Y.C.

1998-01-01

Beryllium-copper reactivity was studied using test parameters being considered for use in the ITER reactor. In this application, beryllium-copper tiles are produced using a low-temperature copper-copper diffusion bonding technique. Beryllium is joined to copper by first plating the beryllium with copper followed by diffusion bonding the electrodeposited (ED) copper to a wrought copper alloy (CuNiBe) at 450 C, 1-3 h using a hot isostatic press (HIP). In this bonded assembly, beryllium is the armor material and the CuNiBe alloy is the heat sink material. Interface temperatures in service are not expected to exceed 350 C. For this study, an ED copper-beryllium interface was subjected to diffusion bonding temperatures and times to study the reaction products. Beryllium-copper assemblies were subjected to 350, 450 and 550 C for times up to 200 h. Both BeCu and Be 2 Cu intermetallic phases were detected using scanning electron microscopy and quantitative microprobe analysis. Growth rates were determined experimentally for each phase and activation energies for formation were calculated. The activation energies were 66 mol and 62 kJ mol -1 for the BeCu and Be 2 Cu, respectively. Tensile bars were produced from assemblies consisting of coated beryllium (both sides) sandwiched between two blocks of Hycon-3. Tensile tests were conducted to evaluate the influence of these intermetallics on the bond strength. Failure occurred at the beryllium-copper interface at fracture strengths greater than 300 MPa for the room-temperature tests. (orig.)

Tritium release from neutron irradiated beryllium pebbles

Energy Technology Data Exchange (ETDEWEB)

Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

1998-01-01

One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

Status of beryllium R and D in Japan

International Nuclear Information System (INIS)

Kawamura, H.; Ishida, K.

2004-01-01

Recently, several R and D program of beryllium for fusion are being promoted in Japan and community of beryllium study is growing up. In the R and D area of beryllium for solid breeding blanket, major subjects are beryllide application for prototype reactor, lifetime evaluation of neutron multiplier, impurity effect of beryllium and recycling of irradiated beryllium. Especially, the study of beryllide application has significant progress in these two years. The basic properties such as tritium inventory, oxidation behavior, steam interaction for stoichiometric Be 12 Ti fabricated by HIP (Hot Isostatic Pressing) have been studied and some advantages against beryllium were made clear. For manufacturing technology development, phase diagram and ductility improvement have been studied. And, Be 12 Ti pebbles with the improved microstructure were successfully fabricated by Rotating Electrode Process. In order to enhance the R and D activities, the R and D network consisted of industries, universities and laboratories in all Japan have been organized. Many collaboration and information exchange strongly promotes the R and D and some projects for commercial application have been launched form these activities. Also international collaborative project such as IEA and ISTC have been launched or planned. Recent result of R and D in Japan is described on this paper. (author)

High conversion heavy water moderated reactor

International Nuclear Information System (INIS)

Miyawaki, Yoshio; Wakabayashi, Toshio.

1989-01-01

In the present invention, fuel rods using uranium-plutonium oxide mixture fuels are arranged in a square lattice at the same pitch as that in light water cooled reactor and heavy water moderators are used. Accordingly, the volume ratio (Vm/Vf) between the moderator and the fuel can be, for example, of about 2. When heavy water is used for the moderator (coolant), since the moderating effect of heavy water is lower than that of light water, a high conversion ratio of not less than 0.8 can be obtained even if the fuel rod arrangement is equal to that of PWR (Vm/Vf about 2). Accordingly, it is possible to avoid problems caused by dense arrangement of fuel rods as in high conversion rate light water cooled reactors. That is, there are no more troubles in view of thermal hydrodynamic characteristics, re-flooding upon loss of coolant accident, etc., as well as the fuel production cost is not increased. (K.M.)

Beryllium. Evaluation of beryllium hydroxide industrial processes. Pt. 3

International Nuclear Information System (INIS)

Lires, O.A.; Delfino, C.A.; Botbol, J.

1991-01-01

This work continues the 'Beryllium' series. It is a historical review of different industrial processes of beryllium hydroxide obtention from beryllium ores. Flowsheats and operative parameters of five plants are provided. These plants (Degussa, Brush Beryllium Co., Beryllium Corp., Murex Ltd., SAPPI) were selected as representative samples of diverse commercial processes in different countries. (Author) [es

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

Refurbishment programme for the BR2-reactor

Energy Technology Data Exchange (ETDEWEB)

Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

1992-07-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Refurbishment programme for the BR2-reactor

International Nuclear Information System (INIS)

Koonen, E.

1992-01-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Compatibility of stainless steels and lithiated ceramics with beryllium

Science.gov (United States)

Flament, T.; Fauvet, P.; Sannier, J.

1988-07-01

The introduction of beryllium as a neutron multiplier in ceramic blankets of thermonuclear fusion reactors may give rise to the following compatibility problems: (i) oxidation of Be by ceramics (lithium aluminate and silicates) or by water vapour; (ii) interaction between beryllium and austenitic and martensitic steels. The studies were done in contact tests under vacuum and in tests under wet sweeping helium. The contact tests under vacuum have revealed that the interaction of beryllium with ceramics seems to be low up to 700°C, the interaction of beryllium with steels is significant and is characterized by the formation of a diffusion layer and of a brittle Be-Fe-Ni compound. With type 316 L austenitic steel, this interaction appears quite large at 600°C whereas it is noticeable only at 700°C with martensitic steels. The experiments carried out with sweeping wet helium at 600°C have evidenced a slight oxidation of beryllium due to water vapour which can be enhanced in the front of uncompletely dehydrated ceramics.

(Beryllium). Internal Report No. 137, Jan. 15, 1958; Le beryllium

Energy Technology Data Exchange (ETDEWEB)

Mouret, P; Rigaud, A

1959-07-01

After a brief summary of the physical and chemical properties of beryllium, the various chemical treatments which can be applied to beryllium minerals either directly or after a physical enrichment are discussed. These various treatments give either the hydroxide or beryllium salts, from which either beryllium oxide or metallic beryllium can easily be obtained. The purification, analysis and uses of beryllium are also briefly discussed. (author)

Beryllium production using beryllium fluoride

International Nuclear Information System (INIS)

Hubler, Carlos Henrique

1993-01-01

This work presents the beryllium production by thermal decomposition of the ammonium beryllium fluoride, followed by magnesium reduction, obtained in the small pilot plant of the Brazilian National Nuclear Energy Commission - Nuclear Engineering Institute

Method of operating heavy water moderated reactors

International Nuclear Information System (INIS)

Masuda, Hiroyuki.

1980-01-01

Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

Heavy water moderated gas-cooled reactors

International Nuclear Information System (INIS)

Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

1964-01-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

Beryllium Project: developing in CDTN of uranium dioxide fuel pellets with addition of beryllium oxide to increase the thermal conductivity

International Nuclear Information System (INIS)

Ferreira, Ricardo Alberto Neto; Camarano, Denise das Merces; Miranda, Odair; Grossi, Pablo Andrade; Andrade, Antonio Santos; Queiroz, Carolinne Mol; Gonzaga, Mariana de Carvalho Leal

2013-01-01

Although the nuclear fuel currently based on pellets of uranium dioxide be very safe and stable, the biggest problem is that this material is not a good conductor of heat. This results in an elevated temperature gradient between the center and its lateral surface, which leads to a premature degradation of the fuel, which restricts the performance of the reactor, being necessary to change the fuel before its full utilization. An increase of only 5 to 10 percent in its thermal conductivity, would be a significant increase. An increase of 50 percent would be a great improvement. A project entitled 'Beryllium Project' was developed in CDTN - Centro de Desenvolvimento da Tecnologia Nuclear, which aimed to develop fuel pellets made from a mixture of uranium dioxide microspheres and beryllium oxide powder to obtain a better heat conductor phase, filling the voids between the microspheres to increase the thermal conductivity of the pellet. Increases in the thermal conductivity in the range of 8.6% to 125%, depending on the level of addition employed in the range of 1% to 14% by weight of beryllium oxide, were obtained. This type of fuel promises to be safer than current fuels, improving the performance of the reactor, in addition to last longer, resulting in great savings. (author)

Investigations of the ternary system beryllium-carbon-tungsten and analyses of beryllium on carbon surfaces; Untersuchung des ternaeren Systems Beryllium-Kohlenstoff-Wolfram und Betrachtungen von Beryllium auf Kohlenstoffoberflaechen

Energy Technology Data Exchange (ETDEWEB)

Kost, Florian

2009-05-25

Beryllium, carbon and tungsten are planned to be used as first wall materials in the future fusion reactor ITER. The aim of this work is a characterization of mixed material formation induced by thermal load. To this end, model systems (layers) were prepared and investigated, which give insight into the basic physical and chemical concepts. Before investigating ternary systems, the first step was to analyze the binary systems Be/C and Be/W (bottom-up approach), where the differences between the substrates PG (pyrolytic graphite) and HOPG (highly oriented pyrolytic graphite) were of special interest. Particularly X-ray photoelectron spectroscopy (XPS), low energy ion scattering (ISS) and Rutherford backscattering spectroscopy (RBS) were used as analysis methods. Beryllium evaporated on carbon shows an island growth mode, whereas a closed layer can be assumed for layer thicknesses above 0.7 nm. Annealing of the Be/C system induces Be{sub 2}C island formation for T{>=}770 K. At high temperatures (T{>=}1170 K), beryllium carbide dissociates, resulting in (metallic) beryllium desorption. For HOPG, carbide formation starts at higher temperatures compared to PG. Activation energies for the diffusion processes were determined by analyzing the decreasing beryllium amount versus annealing time. Surface morphologies were characterized using angle-resolved XPS (ARXPS) and atomic force microscopy (AFM). Experiments were performed to study processes in the Be/W system in the temperature range from 570 to 1270 K. Be{sub 2}W formation starts at 670 K, a complete loss of Be{sub 2}W is observed at 1170 K due to dissociation (and subsequent beryllium desorption). Regarding ternary systems, particularly Be/C/W and C/Be/W were investigated, attaching importance to layer thickness (reservoir) variations. At room temperature, Be{sub 2}C, W{sub 2}C, WC and Be{sub 2}W formation at the respective interfaces was observed. Further Be{sub 2}C is forming with increasing annealing temperatures

Beryllium and copper-beryllium alloys; Beryllium und Kupfer-Beryllium-Legierungen

Energy Technology Data Exchange (ETDEWEB)

Nagel, Nikolaus [Materion Brush GmbH, Stuttgart (Germany). Operation and Quality/EH and S

2017-02-15

The light metal beryllium is a comparatively rare element, which today is primarily derived from bertrandite. It is mainly used as pure metal or in the form of copper-beryllium alloys, e.g., in automotive industry, aerospace, and electrical components. The wide range of applications is mainly attributed to the extremely high rigidity/density ratio. An overview of the history of the metal, its production, and recycling as well as the properties of CuBe alloys are given.

Study on the effect of moderator density reactivity for Kartini reactor

International Nuclear Information System (INIS)

Budi Rohman; Widarto

2009-01-01

One of important characteristics of water-cooled reactors is the change of reactivity due to change in the density of coolant or moderator. This parameter generally has negative value and it has significant role in preventing the excursion of power during operation. Many thermal-hydraulic codes for nuclear reactors require this parameter as the input to account for reactivity feedback due to increase in moderator voids and the subsequent decrease in moderator density during operation. Kartini reactor is cooled and moderated by water, therefore, it is essential to study the effect of the change in moderator density as well as to determine the value of void or moderator density reactivity coefficient in order to characterize its behavior resulting from the presence of vapor or change of moderator density during operation. Analysis by MCNP code shows that the reactivity of core is decreasing with the decrease in moderator density. The analysis estimates the void or moderator density reactivity coefficient for Kartini Reactor to be -2.17×10-4 Δρ/ % void . (author)

Beryllium and zirconium

International Nuclear Information System (INIS)

Salesse, Marc

1959-01-01

Pure beryllium and zirconium, both isolated at about the same date but more than a century ago remained practically unused for eighty years. Fifteen years ago they were released from this state of inactivity by atomic energy, which made them into current metal a with an annual production which runs into tens of tons for the one and thousands for the other. The reasons for this promotion promise well for the future of the two metals, which moreover will probably find additional uses in other branches of industry. The attraction of beryllium and zirconium for atomic energy is easily explained. The curve of figure 1 gives the price per gram of uranium-235 as a function of enrichment: this price increases by about a factor of 3 on passing from natural uranium (0, 7 percent 235 U) to almost pure uranium-235. Because of their tow capture cross-section beryllium and zirconium make it possible, or at least easier, to use natural uranium and they thus enjoy an advantage the extent of which must be calculated for each reactor or fuel element project, but which is generally considerable. It will be seen later that this advantage should be based on figures which are even more favourable that would appear from the simple ratio 3 of the price of pure uranium- 235 contained in natural uranium. Reprint of a paper published in 'Industries Atomiques' - n. 1-2, 1959

Measuring device for the temperature coefficient of reactor moderators

International Nuclear Information System (INIS)

Nakano, Yuzo.

1987-01-01

Purpose: To rapidly determine by automatic calculation the temperature coefficient for moderators which has been determined so far by a log of manual processings. Constitution: Each of signals from a control rod position indicator, a reactor reactivity, instrument and moderator temperature meter are inputted, and each of the signals and designed valued for the doppler temperature coefficients are stored. Recurling calculation is conducted based on the reactivity and the moderator temperature at an interval where the temperature changes of the moderators are equalized at an identical control rod position, to determine isothermic coefficient. Then, the temperature coefficient for moderator are calculated from the isothermic coefficient and the doppler temperature coefficient. The relationship between the reactivity and the moderator temperature is plotted on a X-Y recorder. The stored signals and the calculated temperature coefficient for moderators are sequentially displayed and the results are printed out when the measurement is completed. According to the present device, since the real time processing is conducted, the processing time can be shortened remarkably. Accordingly, it is possible to save the man power for the test of the nuclear reactor and improve the reactor operation performance. (Kamimura, M.)

The high moderating ratio reactor using 100% MOX reloads

International Nuclear Information System (INIS)

Barbrault, P.

1994-06-01

This report presents the concept of a High Moderating ratio Reactor, which should accept 100% MOX reloads. This reactor aims to be the plutonium version of the European Pressurized Reactor (EPR), which is developed jointly by French and German companies. A moderating ration of 2.5 (instead of the standard value of 2.0) is obtained by replacing several fuel rods by water holes. The core would contain 241 Fuel Assemblies. We present some advantages of over-moderation for plutonium fuel, a description of the core and assemblies, calculations of fuel reload schemes and Reactivity Shutdown Margins, and the behavior of the core during two occidental transients. (author). 2 refs., 9 figs., 2 tabs

Status of beryllium study for fusion in RF

International Nuclear Information System (INIS)

Khomutov, A.M.; Kupriyanov, I.B.; Markushkin, Yu.E.; Gervash, A.; Kolbasov, B.N.

2004-01-01

codes have been proposed for determination of the rate of hydrogen generation. The theoretical investigation on the correlation of the plasma disruption with tritium inventory in beryllium components of the first wall of ITER reactor and on the penetration of tritium through first wall material into coolant agents has been performed. Tritium breeding blanket. The investigation on optimization of fabrication processes of porous beryllium with guaranteed predetermined open porosity in the range of 10-30% altogether with properties characterization has been carried out. (author)

Method for welding beryllium

Science.gov (United States)

Dixon, R.D.; Smith, F.M.; O`Leary, R.F.

1997-04-01

A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs.

Method for welding beryllium

International Nuclear Information System (INIS)

Dixon, R.D.; Smith, F.M.; O'Leary, R.F.

1997-01-01

A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs

CFD Simulation on Cooling Down of Beryllium Filters for Neutron Conditioning for Small Angle Neutron Scattering

International Nuclear Information System (INIS)

Azraf Azman; Shahrir Abdullah; Mohd Rizal Mamat

2011-01-01

The cryogenic system for cooling Beryllium filter utilizing liquid nitrogen was designed, fabricated, tested and installed at SANS instrument of TRIGA MARK II PUSPATI research reactor. A computational fluid dynamics (CFD) modeling was used to predict the cooling performance of the beryllium for optimization of neutron beam resolution and transmission. This paper presents the transient CFD results of temperature distributions via the thermal link to the beryllium and simulation of heat flux. The simulation data are also compared with the experimental results for the cooling time and distribution to the beryllium. (author)

Moderator behaviour and reactor internals integrity at Atucha I NPP

International Nuclear Information System (INIS)

Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

1996-01-01

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

Moderator behaviour and reactor internals integrity at Atucha I NPP

Energy Technology Data Exchange (ETDEWEB)

Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

1997-12-31

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

Fast breeder reactor

International Nuclear Information System (INIS)

Ito, Shin-ichi; Maki, Koichi.

1975-01-01

Object: To conserve loaded fuel, aquire controllable surplus reaction degree, increase the breeding index, flatten output and improve sealing of neutrons by inserting a decelerating substance in a blanket section. Structure: A decelerating substance such as beryllium or beryllium oxide is inserted in a blanket section between an outer reactor core and reflector. With this arrangement, neutrons are decelerated to increase the low energy components, which are partly subjected to reflection by the outer reactor core to thereby reduce leakage of neutrons from the reactor core. (Kamimura, M.)

Design and fabrication of a dead weight equipment to perform creep measurements on highly irradiated beryllium specimens

International Nuclear Information System (INIS)

Scibetta, M.; Pellettieri, A.; Wouters, P.; Leenaerts, A.; Verpoucke, G.

2005-01-01

Beryllium is an important material to be used in the blanket of fusion reactors. It acts as a neutron multiplier that allows tritium production. In order to use this material effectively, some data on creep and swelling behaviour are needed. This paper describes preliminary microstructural investigations and the qualification of a creep set-up that will be used to measure creep of highly irradiated beryllium from the BR2 research reactor matrix. (Author)

The Liquid Annular Reactor System (LARS) propulsion

International Nuclear Information System (INIS)

Powell, J.; Ludewig, H.; Horn, F.; Lenard, R.

1990-01-01

A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5)

Reactor-moderated intermediate-energy neutron beams for neutron-capture therapy

International Nuclear Information System (INIS)

Less, T.J.

1987-01-01

One approach to producing an intermediate energy beam is moderating fission neutrons escaping from a reactor core. The objective of this research is to evaluate materials that might produce an intermediate beam for NCT via moderation of fission neutrons. A second objective is to use the more promising moderator material in a preliminary design of an NCT facility at a research reactor. The evaluations showed that several materials or combinations of materials could produce a moderator source for an intermediate beam for NCT. The best neutron spectrum for use in NCT is produced by Al 2 O 3 , but mixtures of Al metal and D 2 O are also attractive. Using the best moderator materials, results were applied to the design of an NCT moderator at the Georgia Institute of Technology Research Reactor's bio-medical facility. The amount of photon shielding and thermal neutron absorber were optimized with respect to the desired photon dose rate and intermediate neutron flux at the patient position

Fluid moderator control system reactor internals distribution system

International Nuclear Information System (INIS)

Fensterer, H.F.; Klassen, W.E.; Veronesi, L.; Boyle, D.E.; Salton, R.B.

1987-01-01

This patent describes a spectral shift pressurized water nuclear reactor employing a low neutron moderating fluid for the spectral shift including a reactor pressure vessel, a core comprising a plurality of fuel assemblies, a core support plate, apparatus comprising means for penetrating the reactor vessel for introducing the moderating fluid into the reactor vessel. Means associated with the core support plate for directly distributing the moderating fluid to and from the fuel assemblies comprises at least one inlet flow channel in the core plate; branch inlet feed lines connect to the inlet flow channel in the core plate; vertical inlet flow lines flow connected to the branch inlet feed lines; each vertical flow line communicates with a fuel assembly; the distribution means further comprise lines serving as return flow lines, each of which is connected to one of the fuel assemblies; branch exit flow lines in the core plate flow connected to the return flow lines of the fuel assembly; and at least one outlet flow channel flow connected to the branch exit flow lines; and a flow port interposed between the penetration means and the distribution means for flow connecting the penetration means with the distribution means

Fusion Reactor Materials

International Nuclear Information System (INIS)

Moons, F.

1998-01-01

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

Beryllium irradiation embrittlement test programme. Material and specimen specification, manufacture and qualification

International Nuclear Information System (INIS)

Harries, D.R.; Dalle Donne, M.

1996-06-01

The report presents the specification, manufacture and qualification of the beryllium specimens to be irradiated in the BR2 reactor in Mol to investigate the effect of the neutron irradiation on the embrittlement as a function of temperature and beryllium oxide content. This work was been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Union within the European Fusion Technology Program. (orig.)

Investigations of the ternary system beryllium-carbon-tungsten and analyses of beryllium on carbon surfaces

International Nuclear Information System (INIS)

Kost, Florian

2009-01-01

Beryllium, carbon and tungsten are planned to be used as first wall materials in the future fusion reactor ITER. The aim of this work is a characterization of mixed material formation induced by thermal load. To this end, model systems (layers) were prepared and investigated, which give insight into the basic physical and chemical concepts. Before investigating ternary systems, the first step was to analyze the binary systems Be/C and Be/W (bottom-up approach), where the differences between the substrates PG (pyrolytic graphite) and HOPG (highly oriented pyrolytic graphite) were of special interest. Particularly X-ray photoelectron spectroscopy (XPS), low energy ion scattering (ISS) and Rutherford backscattering spectroscopy (RBS) were used as analysis methods. Beryllium evaporated on carbon shows an island growth mode, whereas a closed layer can be assumed for layer thicknesses above 0.7 nm. Annealing of the Be/C system induces Be 2 C island formation for T≥770 K. At high temperatures (T≥1170 K), beryllium carbide dissociates, resulting in (metallic) beryllium desorption. For HOPG, carbide formation starts at higher temperatures compared to PG. Activation energies for the diffusion processes were determined by analyzing the decreasing beryllium amount versus annealing time. Surface morphologies were characterized using angle-resolved XPS (ARXPS) and atomic force microscopy (AFM). Experiments were performed to study processes in the Be/W system in the temperature range from 570 to 1270 K. Be 2 W formation starts at 670 K, a complete loss of Be 2 W is observed at 1170 K due to dissociation (and subsequent beryllium desorption). Regarding ternary systems, particularly Be/C/W and C/Be/W were investigated, attaching importance to layer thickness (reservoir) variations. At room temperature, Be 2 C, W 2 C, WC and Be 2 W formation at the respective interfaces was observed. Further Be 2 C is forming with increasing annealing temperatures. Depending on the layer

Comparative thermal cyclic test of different beryllium grades previously subjected to simulated disruption loads

International Nuclear Information System (INIS)

Gervash, A.; Giniyatulin, R.; Mazul, I.

1999-01-01

Considering beryllium as plasma facing armour this paper presents recent results obtained in Russia. A special process of joining beryllium to a Cu-alloy material structure is described and recent results of thermal cycling tests of such joints are presented. Summarizing the results, the authors show that a Cu-alloy heat sink structure armoured with beryllium can survive high heat fluxes (≥10 MW/m 2 ) during 1000 heating/cooling cycles without serious damage to the armour material and its joint. The principal feasibility of thermal cycling of beryllium grades and their joints directly in the core of a nuclear reactor is demonstrated and the main results of this test are presented. The paper also describes the thermal cycling of different beryllium grades having cracks initiated by previously applied high heat loads simulating plasma disruptions. (orig.)

Beryllium and growth. II. The effect of beryllium on plant growth

Energy Technology Data Exchange (ETDEWEB)

Hoagland, M B

1952-01-01

Experiments were undertaken to determine whether beryllium could replace magnesium in a growing organism. This was stimulated by the several known growth effects of beryllium in animals and by the fact that beryllium apparently competes with magnesium for animal alkaline phosphatases. The following findings are noted: (1) beryllium can reduce the magnesium requirement of plants by some 60% within a certain range of magnesium deficiency. (2) The residual obligatory magnesium requirements is probably accounted for by chlorophyll since beryllium appears to have no primary effect on chlorophyll or chlorophyll production. (3) The pH of the nutrient solution is critical: at acid pH's, beryllium is highly toxic, and growth increase due to beryllium only appears at initial pH's above 11.2, although this initial pH rapidly falls to neutrality during the experimental period. 22 references, 4 figures, 1 table.

Status and perspective of development of cold moderators at the IBR-2 reactor

International Nuclear Information System (INIS)

Kulikov, S; Shabalin, E

2012-01-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams nos. 7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams nos. 2-3 and for beams nos. 1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (∼3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

Status and perspective of development of cold moderators at the IBR-2 reactor

Science.gov (United States)

Kulikov, S.; Shabalin, E.

2012-03-01

The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

Plasma spraying of beryllium and beryllium-aluminum-silver alloys

International Nuclear Information System (INIS)

Castro, R.G.; Stanek, P.W.; Elliott, K.E.; Jacobson, L.A.

1994-01-01

A preliminary investigation on plasma-spraying of beryllium and a beryllium-aluminum-4% silver alloy was done at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility (BATSF). Spherical Be and Be-Al-4%Ag powders, which were produced by centrifugal atomization, were used as feedstock material for plasma-spraying. The spherical morphology of the powders allowed for better feeding of fine (<38 μm) powders into the plasma-spray torch. The difference in the as-deposited densities and deposit efficiencies of the two plasma-sprayed powders will be discussed along with the effect of processing parameters on the as-deposited microstructure of the Be-Al-4%Ag. This investigation represents ongoing research to develop and characterize plasma-spraying of beryllium and beryllium-aluminum alloys for magnetic fusion and aerospace applications

Plasma spraying of beryllium and beryllium-aluminum-silver alloys

International Nuclear Information System (INIS)

Castro, R.G.; Stanek, P.W.; Elliott, K.E.; Jacobson, L.A.

1993-01-01

A preliminary investigation on plasma-spraying of beryllium and a beryllium-aluminum 4% silver alloy was done at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility (BATSF). Spherical Be and Be-Al-4%Ag powders, which were produced by centrifugal atomization, were used as feedstock material for plasma-spraying. The spherical morphology of the powders allowed for better feeding of fine (<38 μm) powders into the plasma-spray torch. The difference in the as-deposited densities and deposit efficiencies of the two plasma-sprayed powders will be discussed along with the effect of processing parameters on the as-deposited microstructure of the Be-Al-4%Ag. This investigation represents ongoing research to develop and characterize plasma-spraying of beryllium and beryllium-aluminum alloys for magnetic fusion and aerospace applications

Experimental and thermodynamic studies of beryllium replacement materials for CANDU brazed joints

Energy Technology Data Exchange (ETDEWEB)

Potter, K.N.; Ferrier, G.A.; Corcoran, E.C., E-mail: Kieran.Potter@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

2015-07-01

Currently, appendages are joined to CANDU fuel elements via a brazing process, which uses beryllium as the filler material. A potential reduction in the occupational limit on airborne beryllium particulates has motivated research into alternative brazing materials. To this end, the Canadian nuclear industry has funded an initiative to identify and evaluate the suitability of several candidate materials. This work describes contributions toward the assessment of alternative brazing materials from the Royal Military College of Canada. Thermodynamic modelling was performed to predict the aqueous behaviour of each candidate material in CANDU coolant conditions characteristic of reactor shutdown, and experiments are underway to support modelling predictions. These results will assist in selecting a suitable replacement material for beryllium. (author)

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

Summary of the 4th workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

The JFS libraries and the effect of beryllium on the breeding

International Nuclear Information System (INIS)

Ishiguro, Y.; Gouveia, A.S. de.

1981-02-01

The Japanese group constants libraries for fast reactor analysis JFS-25 and JFS-70 are compared. The effect of beryllium on the breeding characteristics of thorium cycle breeders is examined. The results show that the 25-group set predicts shorter reactor doubling times than the 70-group set and that the effect of Be(n,2n) reactions is negligible. (Author) [pt

Effects of moderation level on core reactivity and. neutron fluxes in natural uranium fueled and heavy water moderated reactors

International Nuclear Information System (INIS)

Khan, M.J.; Aslam; Ahmad, N.; Ahmed, R.; Ahmad, S.I.

2005-01-01

The neutron moderation level in a nuclear reactor has a strong influence on core multiplication, reactivity control, fuel burnup, neutron fluxes etc. In the study presented in this article, the effects of neutron moderation level on core reactivity and neutron fluxes in a typical heavy water moderated nuclear research reactor is explored and the results are discussed. (author)

Influence of impurities in Beryllium on tritium breeding ratio

International Nuclear Information System (INIS)

Yamauchi, M.; Ochiai, K.; Verzilov, Y.; Ito, M.; Wada, M.; Nishitani, T.

2004-01-01

Several neutronics experiments simulating fusion blankets have been conducted with 14 MeV neutron source to assess the reliability of nuclear analysis codes. However, the analyses have not always presented good agreements so far between calculated and measured tritium production rates. One of the reasons was considered as impurities in beryllium which has negligibly small neutron absorption cross section in low energy range. Chemical compositions of beryllium were analyzed by Inductively Coupled Plasma (ICP) method, and a pulsed neutron decay experiment discovered that the macroscopic neutron absorption cross section for beryllium medium may be about 30% larger than the value calculated by the data specified by manufacturing company. The influence of the impurities on the calculations was studied on the basis of the fusion DEMO-reactor blanket design. As a result of the study, it was made clear that the impurities affect the local tritium production rates when the size of beryllium medium is more than 20-30 mean free paths (30-40 cm) in thickness. In case of some blanket designs that meet the above condition, the effect on tritium breeding ratio may become as large as about 4%. (author)

Influence of impurities in Beryllium on tritium breeding ratio

Energy Technology Data Exchange (ETDEWEB)

Yamauchi, M; Ochiai, K; Verzilov, Y; Ito, M; Wada, M; Nishitani, T [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2004-03-01

Several neutronics experiments simulating fusion blankets have been conducted with 14 MeV neutron source to assess the reliability of nuclear analysis codes. However, the analyses have not always presented good agreements so far between calculated and measured tritium production rates. One of the reasons was considered as impurities in beryllium which has negligibly small neutron absorption cross section in low energy range. Chemical compositions of beryllium were analyzed by Inductively Coupled Plasma (ICP) method, and a pulsed neutron decay experiment discovered that the macroscopic neutron absorption cross section for beryllium medium may be about 30% larger than the value calculated by the data specified by manufacturing company. The influence of the impurities on the calculations was studied on the basis of the fusion DEMO-reactor blanket design. As a result of the study, it was made clear that the impurities affect the local tritium production rates when the size of beryllium medium is more than 20-30 mean free paths (30-40 cm) in thickness. In case of some blanket designs that meet the above condition, the effect on tritium breeding ratio may become as large as about 4%. (author)

Failure analysis of beryllium tile assembles following high heat flux testing for the ITER program

International Nuclear Information System (INIS)

Odegard, B.C. Jr.; Cadden, C. H.; Yang, N. Y. C.

2000-01-01

The following document describes the processing, testing and post-test analysis of two Be-Cu assemblies that have successfully met the heat load requirements for the first wall and dome sections for the ITER (International Thermonuclear Experimental Reactor) fusion reactor. Several different joint assemblies were evaluated in support of a manufacturing technology investigation aimed at diffusion bonding or brazing a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Judicious selection of materials and coatings for these assemblies was essential to eliminate or minimize interactions with the highly reactive beryllium armor material. A thin titanium layer was used as a diffusion barrier to isolate the copper heat sink from the beryllium armor. To reduce residual stresses produced by differences in the expansion coefficients between the beryllium and copper, a compliant layer of aluminum or aluminum-beryllium (AlBeMet-150) was used. Aluminum was chosen because it does not chemically react with, and exhibits limited volubility in, beryllium. Two bonding processes were used to produce the assemblies. The primary process was a diffusion bonding technique. In this case, undesirable metallurgical reactions were minimized by keeping the materials in a solid state throughout the fabrication cycle. The other process employed an aluminum-silicon layer as a brazing filler material. In both cases, a hot isostatic press (HIP) furnace was used in conjunction with vacuum-canned assemblies in order to minimize oxidation and provide sufficient pressure on the assemblies for full metal-to-metal contact and subsequent bonding. The two final assemblies were subjected to a suite of tests including: tensile tests and electron and optical metallography. Finally, high heat flux testing was conducted at the electron beam testing system (EBTS) at Sandia National Laboratories, New Mexico. Here, test mockups were fabricated and subjected to normal heat loads to

Beryllium

International Nuclear Information System (INIS)

Hansen, N.B.

1980-01-01

A method for determination of beryllium in minerals and rocks is described. The method comprises microanalysis and trace analysis. Because of the toxidity of beryllium the method is designed for determination of a hitherto unknown small amount, 1-10 nanogram Be. With the optimal amount for determination, 3 ng Be, the relative error of the method is 10%. The description includes an inventory of chemicals and apparatus, also an example of application of the method on the mineral epididymite. In brief, the sample is melted with sodium carbonate and sodium tetra borate; when required the sample in advance is fumed with hydrogen fluoride and sulphuric acid to evaporate silica. The residuum is dissolved in water and hydrogen chloride, upon which the solution is made to volume. In the Ring oven interfering compounds are masked with EDTA. Beryllium is settled with chrome azurol and ammonia. Beryllium is identified and evaluated in comparison with previously produced standards. (author)

Beryllium

International Nuclear Information System (INIS)

Hansen, N.B.

1979-01-01

A method for determination of beryllium in minerals and rocks is described. The method comprises microanalysis and trace analysis. Because of the toxidity of beryllium the method is designed for determination of a hitherto unknown small amount, 1-10 nanogram Be. With the optimal amount for determination, 3 ng Be, the relative error of the method is 10%. The description includes an inventory of chemicals and apparatus, also an example of application of the method on the mineral epididymite. In brief, the sample is melted with sodium carbonate and sodium tetra borate; when required the sample in advance is fumed with hydrogen fluoride and sulphuric acid to evaporate silica. The residuum is dissolved in water and hydrogen chloride, upon which the solution is made to volume. In the Ring oven interfering compounds are masked with EDTA. Beryllium is settled with chrome azurol and ammonia. Beryllium is identified and evaluated in comparison with previously produced standards. (author)

On modeling of beryllium molten depths in simulated plasma disruptions

International Nuclear Information System (INIS)

Tsotridis, G.; Rother, H.

1996-01-01

Plasma-facing components in tokamak-type fusion reactors are subjected to intense heat loads during plasma disruptions. The influence of high heat fluxes on the depth of heat-affected zones of pure beryllium metal and beryllium containing very low levels of surface active impurities is studied by using a two-dimensional transient computer model that solves the equations of motion and energy. Results are presented for a range of energy densities and disruption times. Under certain conditions, impurities, through their effect on surface tension, create convective flows and hence influence the flow intensities and the resulting depths of the beryllium molten layers during plasma disruptions. The calculated depths of the molten layers are also compared with other mathematical models that are based on the assumption that heat is transported through the material by conduction only. 32 refs., 6 figs., 1 tab

Beryllium health effects in the era of the beryllium lymphocyte proliferation test.

Science.gov (United States)

Maier, L A

2001-05-01

The beryllium lymphocyte proliferation test (BeLPT) has revolutionized our approach to the diagnosis, screening, and surveillance of beryllium health effects. Based on the development of a beryllium-specific cell-mediated immune response, the BeLPT has allowed us to define early health effects of beryllium, including beryllium sensitization (BeS), and chronic beryllium disease (CBD) at a subclinical stage. The use of this test as a screening tool has improved our understanding of these health effects. From a number of studies it is apparent that BeS precedes CBD and develops after as little as 9 weeks of beryllium exposure. CBD occurs within 3 months and up to 30 years after initial beryllium exposure. Exposure-response variables have been associated with BeS/CBD, including work as a machinist, chemical or metallurgical operator, laboratory technician, work in ceramics or beryllium metal production, and years of beryllium exposure. Recent studies have found BeS and CBD in workplaces in which the majority of exposures were below the 2 microg/m3 OSHA time-weighted average (TWA). Ideally, the BeLPT would be used in surveillance aimed at defining other risk-related processes, determining exposure variables which predict BeS and CBD, and defining the exposure level below which beryllium health effects do not occur. Unfortunately, the BeLPT can result in false negative tests and still requires an invasive procedure, a bronchoscopy, for the definitive diagnosis of CBD. Thus, research is needed to establish new tests to be used alone or in conjunction with the BeLPT to improve our ability to detect early beryllium health effects.

Reactivity variations associated with the core expansion of the MARIA research reactor after modernisation

International Nuclear Information System (INIS)

Krzysztoszek, G.

1997-01-01

Polish high flux research reactor MARIA is a pool type reactor moderated with beryllium and water and cooled with water. The fuel is 80% enriched uranium, in the shape of multitube fuel elements, each tube made up of UAl x alloy in aluminium cladding. MARIA reactor has been operated in the years of 1977-85 and then it was modernised and again put into operation in December 1992. The modernisation as regarded the reactor core comprises a beryllium matrix expansion from 20-48 blocks. Within the frame of the power start-up and trial operation the reactor has been extended from 12 to 18 fuel channels. On that stage of reactor operation the power of mostly loaded fuel channels was constrained to 1,6 MW. Reactor has been operated within the 100-hrs campaign for an irradiation of target materials and for performing measurements at the horizontal channel outlets. In the previous time it has been noticed substantial differences in reactivity changes of the core in similar campaigns of reactor operation. It concerns the reactivity losses during poisoning period of the reactor within the first 30-40 hrs of operation as well as in the fuel burning up process. An analysis of the reactivity variations during the core extension will made possible the fuel management optimisation in further reactor operation system. (author)

Postirradiation examination of beryllium pebbles

International Nuclear Information System (INIS)

Gelles, D.S.

1998-01-01

Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements

Assessment of the feasibility and advantages of beryllium recycling

International Nuclear Information System (INIS)

Druyts, F.; Braet, J.; Ooms, L.

2006-01-01

This paper proposes a generic route for the recycling of beryllium from fusion reactors, based on critical issues associated with beryllium pebbles after their service life in the HCPB breeding blanket. These critical issues are the high tritium inventory, the presence of long-lived radionuclides (among which transuranics due to traces of uranium in the base metal), and the chemical toxicity of beryllium. On the basis of the chemical and radiochemical characteristics of the neutron irradiated beryllium pebbles, we describe a possible recycling route. The first step is the detritiation of the material. This can be achieved by heating the pebbles to 800 o C under an argon flow. The argon gas avoids oxidation of the beryllium, and at the proposed temperature the tritium inventory is readily released from the pebbles. In a second step, the released tritium can be oxidised on a copper oxide bed to produce tritiated water, which is consistent with the current international strategy to convert all kinds of tritiated waste into tritiated water, which can subsequently be treated in a water detritiation plant. Removal of radionuclides from the beryllium pebbles may be achieved by several types of chloride processes. The first step is to pass chlorine gas (in an argon flow) over the pebbles, thus yielding volatile BeCl 2 . This beryllium chloride can then be purified by fractional distillation. As a small fraction of the beryllium chloride contains the long-lived 10 Be isotope, 10BeCl 2 has to be separated from 9BeCl 2 , which could be achieved by centrifugal techniques. The product can then be reduced to obtain high-purity metallic beryllium. Two candidate reduction methods were identified: fused salt electrolysis and thermal decomposition. Both these methods require laboratory parametric studies to maximise the yield and achieve a high purity metal, before either process can be upgraded to a larger scale. The eventual product of the chloride reduction process must be a high

Summary of the 4th workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi (eds.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-09-01

The research on Reduced-Moderation Water Reactors (RMWRs) has been performed in JAERI for the development of future innovative reactors. The workshop on the RMWRs has been held every year since fiscal 1997 aimed at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors. The 4th workshop was held on March 2, 2001 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The workshop began with three lectures on recent research activities in JAERI entitled 'Recent Situation of Research on Reduced-Moderation Water Reactor', 'Analysis on Electricity Generation Costs of Reduced Moderation Water Reactors' and 'Reprocessing Technology for Spent Mixed-Oxides Fuel from LWR'. Then five lectures followed: 'Micro Reactor Physics of MOX Fueled LWR' which shows the recent results of reactor physics, Fast Reactor Cooled by Supercritical Light Water' which is another type of reduced-moderation reactor, 'Phase 1 of Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC), 'Integral Type Small PWR with Stand-alone Safety' which is intended to suit for the future consumers' needs, and Utilization of Plutonium in Reduced-Moderation Water Reactors' which dictates benefits of plutonium utilization with RMWRs. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as presentation handouts, program and participant list as appendixes. The 8 of the presented papers are indexed individually. (J.P.N.)

Reactivity costs in MARIA reactor

International Nuclear Information System (INIS)

Marcinkowska, Zuzanna E.; Pytel, Krzysztof M.; Frydrysiak, Andrzej

2017-01-01

Highlights: • The methodology for calculating consumed fuel cost of excess reactivity is proposed. • Correlation between time integral of the core excess reactivity and released energy. • Reactivity price gives number of fuel elements required for given excess reactivity. - Abstract: For the reactor operation at high power level and carrying out experiments and irradiations the major cost of reactor operation is the expense of nuclear fuel. In this paper the methodology for calculating consumed fuel cost-relatedness of excess reactivity is proposed. Reactivity costs have been determined on the basis of operating data. A number of examples of calculating the reactivity costs for processes such as: strong absorbing material irradiation, molybdenium-99 production, beryllium matrix poisoning and increased moderator temperature illustrates proposed method.

Laser fabrication of beryllium components

International Nuclear Information System (INIS)

Hanafee, J.E.; Ramos, T.J.

1995-08-01

Working with the beryllium industry on commercial applications and using prototype parts, the authors have found that the use of lasers provides a high-speed, low-cost method of cutting beryllium metal, beryllium alloys, and beryllium-beryllium oxide composites. In addition, they have developed laser welding processes for commercial structural grades of beryllium that do not need a filler metal; i.e., autogenous welds were made in commercial structural grades of beryllium by using lasers

First-principles and classical molecular dynamics study of threshold displacement energy in beryllium

Energy Technology Data Exchange (ETDEWEB)

Vladimirov, P.V. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany); Borodin, V.A., E-mail: Borodin_VA@nrcki.ru [National Research Center “Kurchatov Institute”, 123182 Moscow (Russian Federation); NRNU MEPhI, 115409 Moscow (Russian Federation)

2017-02-15

Highlights: • Beryllium is a functional material of future fusion reactors. • The threshold displacement energy by fast particles is studied. • Classical and first principles simulations are used. - Abstract: Beryllium selected as a neutron multiplier material for the tritium breeding blanket of fusion reactor should withstand high doses of fast neutron irradiation. The damage produced by irradiation is usually evaluated assuming that the number of atomic displacements to the threshold displacement energy, E{sub d}, which is considered as an intrinsic material parameter. In this work the value of E{sub d} for hcp beryllium is estimated simultaneously from classical and first-principles molecular dynamics simulations. Quite similar quantitative pictures of defect production are observed in both simulation types, though the predicted displacement threshold values seem to be approximately two times higher in the first-principles approach. We expect that, after more detailed first-principles investigations, this approach can be used for scaling the damage prediction predictions by classical molecular dynamics, opening a way for more consistent calculations of displacement damage in materials.

Permeation behavior of deuterium implanted into beryllium

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hirofumi; Hayashi, Takumi; O' hira, Shigeru; Nishi, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-09-01

Study on Implantation Driven Permeation (IDP) behavior of deuterium through pure beryllium was investigated as a part of the research to predict the tritium permeation through the first wall components ITER (International Thermonuclear Experimental Reactor). The permeation experiments were carried out with two beryllium specimens, one was an unannealed specimen and the other was that annealed at 1173 K. The permeation flux was measured as a function of specimen temperature and incident ion flux. Surface analysis of specimen was also carried out after the permeation experiment. Permeation was observed only with the annealed specimen and no significant permeation was observed with unannealed specimen under the present experimental condition (maximum temperature: 685 K, detection limit: 1x10{sup 13} D atoms/m{sup 2}s). It could be attributed that the intrinsic lattice defects, which act as diffusion preventing site, decreased with the specimen annealing. Based on the result of steady and transient permeation behavior and surface analysis, it was estimated that the deuterium permeation implanted into annealed beryllium was controlled by surface recombination due to the oxide layer on the surface of the permeated side. (author)

Beryllium and growth. III. The effect of beryllium on plant phosphatase

Energy Technology Data Exchange (ETDEWEB)

Hoagland, M B

1952-01-01

The purpose of the investigations was to correlate the apparent ability of beryllium to substitute for magnesium in plant growth with a specific biochemical effect of the metal. Through association with earlier work on beryllium inhibition of animal alkaline phosphatase, a study was made of the effect of beryllium and other metals upon the activity of a phosphatase derived from tomato leaves. Although only indirect evidence is available that this enzyme system was magnesium-activated, beryllium was found to inhibit reversibly the splitting of GP and ATP. Other metals were also found to be inhibitory but the ATP-ase inhibition - and especially the ratio of P split from GP to P split from ATP - was higher for beryllium than for any other metal studied. The significance of this finding in relation to energy metabolism, growth, and beryllium toxicity is discussed. 12 references, 5 figures, 2 tables.

Graphite-moderated and heavy water-moderated spectral shift controlled reactors

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs

Beryllium chemistry and processing

CERN Document Server

Walsh, Kenneth A

2009-01-01

This book introduces beryllium; its history, its chemical, mechanical, and physical properties including nuclear properties. The 29 chapters include the mineralogy of beryllium and the preferred global sources of ore bodies. The identification and specifics of the industrial metallurgical processes used to form oxide from the ore and then metal from the oxide are thoroughly described. The special features of beryllium chemistry are introduced, including analytical chemical practices. Beryllium compounds of industrial interest are identified and discussed. Alloying, casting, powder processing, forming, metal removal, joining and other manufacturing processes are covered. The effect of composition and process on the mechanical and physical properties of beryllium alloys assists the reader in material selection. The physical metallurgy chapter brings conformity between chemical and physical metallurgical processing of beryllium, metal, alloys, and compounds. The environmental degradation of beryllium and its all...

LIGHT WATER MODERATED NEUTRONIC REACTOR

Science.gov (United States)

Christy, R.F.; Weinberg, A.M.

1957-09-17

A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

Beryllium. Beryllium oxide, obtention and properties. Pt.4

International Nuclear Information System (INIS)

Lires, O.A.; Delfino, C.A.; Botbol, J.

1991-01-01

As a continuation of the 'Beryllium' series this work reviews several methods of high purity beryllia production. Diverse methods of obtention and purification from different beryllium compounds are described. Some chemical, mechanical and electrical properties related with beryllia obtention methods are summarized. (Author) [es

Beryllium

International Nuclear Information System (INIS)

1988-01-01

In this data sheet the occurrence, ore processing, chemical and physical properties and the uses of beryllium and its alloys is presented. The hazards involved in the use of beryllium and its compounds in the laboratory are discussed with particular reference to its toxicity, carcinogenicity, handling, storage, disposal, fire prevention and the principal health hazards. Further reading is suggested. (UK)

Solid methane cold moderator for the IBR-2 reactor

International Nuclear Information System (INIS)

Beliakov, A. A.; Tretiakov, I. T.; Shabalin, E. P.; Golikov, V. V.; Luschivkov, V I.

1997-09-01

The paper describes the research and design work carried out since 1986 at the Frank Laboratory of Neutron Physics of the Joint Institute for Nuclear Research in Dubna to create a cryogenic moderator for the IBR-22 reactor using solid methane as a moderating substance.

Hanford Site Beryllium Program: Past, Present, and Future - 12428

Energy Technology Data Exchange (ETDEWEB)

Fisher, Mark [CH2M Hill Plateau Remediation Company, Richland, Washington 99354 (United States); Garcia, Pete [U.S. Department of Energy - Richland Office, Richland, Washington 99352 (United States); Goeckner, Julie [U.S. Department of Energy - HQ, EMCBC, Cincinnati, Ohio 45202 (United States); Millikin, Emily [Washington Closure Hanford, Richland, Washington 99354 (United States); Stoner, Mike [Mission Support Alliance, Richland, Washington 99354 (United States)

2012-07-01

The U.S. Department of Energy (DOE) has a long history of beryllium use because of the element's broad application to many nuclear operations and processes. At the Hanford Site beryllium alloy was used to fabricate parts for reactors, including fuel rods for the N-Reactor during plutonium production. Because of continued confirmed cases of chronic beryllium disease (CBD), and data suggesting CBD occurs at exposures to low-level concentrations, the DOE decided to issue a rule to further protect federal and contractor workers from hazards associated with exposure to beryllium. When the beryllium rule was issued in 1999, each of the Hanford Site contractors developed a Chronic Beryllium Disease Prevention Program (CBDPP) and initial site wide beryllium inventories. A new site-wide CBDPP, applicable to all Hanford contractors, was issued in May, 2009. In the spring of 2010 the DOE Headquarters Office of Health, Safety, and Security (HSS) conducted an independent inspection to evaluate the status of implementation of the Hanford Site Chronic Beryllium Disease Prevention Program (CBDPP). The report identified four Findings and 12 cross-cutting Opportunities for Improvement (OFIs). A corrective action plan (CAP) was developed to address the Findings and crosscutting OFIs. The DOE directed affected site contractors to identify dedicated resources to participate in development of the CAP, along with involving stakeholders. The CAP included general and contractor-specific recommendations. Following initiation of actions to implement the approved CAP, it became apparent that additional definition of product deliverables was necessary to assure that expectations were adequately addressed and CAP actions could be closed. Consequently, a supplement to the original CAP was prepared and transmitted to DOE-HQ for approval. Development of the supplemental CAP was an eight month effort. From the onset a core group of CAP development members were identified to develop a mechanism

Extractive metallurgy of the beryllium

International Nuclear Information System (INIS)

Alonso, Neusa; Capocchi, Jose Deodoro Trani

1995-01-01

A bibliographic review is performed on the beryllium extractive metallurgy. The work describes the main type of ores and processes applied to the metallic beryllium production, beryllium oxide production using fluoride, sulfide and direct chlorination. The thermodynamic consideration are made on beryllium reduction processes, discussing the viability of the beryllium oxide and hallide reduction processes. Under the technological viewpoint, the Cu-Be alloys main production processes are discussed, and the main toxicity problems related with beryllium are mentioned

FLUID MODERATED REACTOR

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

1957-10-22

A reactor which utilizes fissionable fuel elements in rod form immersed in a moderator or heavy water and a means of circulating the heavy water so that it may also function as a coolant to remove the heat generated by the fission of the fuel are described. In this design, the clad fuel elements are held in vertical tubes immersed in heavy water in a tank. The water is circulated in a closed system by entering near the tops of the tubes, passing downward through the tubes over the fuel elements and out into the tank, where it is drawn off at the bottom, passed through heat exchangers to give up its heat and then returned to the tops of the tubes for recirculation.

Technical Basis for PNNL Beryllium Inventory

Energy Technology Data Exchange (ETDEWEB)

Johnson, Michelle Lynn

2014-07-09

The Department of Energy (DOE) issued Title 10 of the Code of Federal Regulations Part 850, “Chronic Beryllium Disease Prevention Program” (the Beryllium Rule) in 1999 and required full compliance by no later than January 7, 2002. The Beryllium Rule requires the development of a baseline beryllium inventory of the locations of beryllium operations and other locations of potential beryllium contamination at DOE facilities. The baseline beryllium inventory is also required to identify workers exposed or potentially exposed to beryllium at those locations. Prior to DOE issuing 10 CFR 850, Pacific Northwest Nuclear Laboratory (PNNL) had documented the beryllium characterization and worker exposure potential for multiple facilities in compliance with DOE’s 1997 Notice 440.1, “Interim Chronic Beryllium Disease.” After DOE’s issuance of 10 CFR 850, PNNL developed an implementation plan to be compliant by 2002. In 2014, an internal self-assessment (ITS #E-00748) of PNNL’s Chronic Beryllium Disease Prevention Program (CBDPP) identified several deficiencies. One deficiency is that the technical basis for establishing the baseline beryllium inventory when the Beryllium Rule was implemented was either not documented or not retrievable. In addition, the beryllium inventory itself had not been adequately documented and maintained since PNNL established its own CBDPP, separate from Hanford Site’s program. This document reconstructs PNNL’s baseline beryllium inventory as it would have existed when it achieved compliance with the Beryllium Rule in 2001 and provides the technical basis for the baseline beryllium inventory.

Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

International Nuclear Information System (INIS)

Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

2017-01-01

Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

Summary of the 3rd workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Summary of the 3rd workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Spectrographic measurement of beryllium in the atmosphere; Dosage spectrographique du beryllium dans l'atmosphere

Energy Technology Data Exchange (ETDEWEB)

Artaud, J; Cittanova, J [Commissariat a l' Energie Atomique, Service d' Analyses et Recherches Chimiques Appliquees, Saclay (France). Centre d' Etudes Nucleaires; Crehange, G; Frequelin, S [Commissariat a l' Energie Atomique, Dir. des Applications Militaires, Service Chimie, Saclay (France). Centre d' Etudes Nucleaires; Baudin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1961-07-01

We describe here a method for the spectrographic determination of beryllium on filters which is valid for amounts varying between 0,01 and 30 {mu}g of beryllium and which is independent of the nature of the beryllium compound involved. This is a flux method (graphite-lithium carbonate mixture), the excitation being by a direct current arc. (author) [French] Nous decrivons ici, une methode de dosage spectrographique de beryllium sur filtre, valable pour des teneurs comprises entre 0,01 et 30 {mu}g de beryllium et independante de la nature du compose de beryllium a doser. C'est une methode de 'flux' (melange graphite-carbonate de lithium) l'excitation etant un arc a courant continu. (auteur)

Experimental and thermodynamic assessment of beryllium-replacement materials for CANDU brazed joints

Energy Technology Data Exchange (ETDEWEB)

Potter, K.N.; Ferrier, G.A.; Corcoran, E.C., E-mail: Kieran.Potter@rmc.ca [Royal Military College of Canada, Kingston ON, (Canada); Dimayuga, F.C. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

2015-07-01

Currently, appendages are joined to CANDU fuel elements via a brazing process, with beryllium as the filler material. A potential reduction in the occupational limit on airborne beryllium particulates has motivated research into alternative brazing materials. To this end, the Canadian nuclear industry has funded an initiative to identify and evaluate the suitability of several candidate brazing materials. This work describes contributions toward the assessment of alternative brazing materials from the Royal Military College of Canada (RMCC). An impact testing method was developed to evaluate the mechanical strength of candidate braze joints.Thermodynamic modelling was performed to predict the aqueous behaviour of each candidate material in CANDU coolant conditions characteristic of reactor shutdown, and corrosion experiments are underway to support modelling predictions.The results of these activities will assist in selecting a suitable replacement material for beryllium. (author)

Tritium release and retention properties of highly neutron-irradiated beryllium pebbles from HIDOBE-01 experiment

Energy Technology Data Exchange (ETDEWEB)

Chakin, V., E-mail: vladimir.chakin@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R.; Moeslang, A.; Klimenkov, M.; Kolb, M.; Vladimirov, P.; Kurinskiy, P.; Schneider, H.-C. [Karlsruhe Institute of Technology, Institute for Applied Materials, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Til, S. van; Magielsen, A.J. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

2013-11-15

The current helium cooled pebble bed (HCPB) tritium breeding blanket concept for fusion reactors includes a bed of 1 mm diameter beryllium pebbles to act as a neutron multiplier. Beryllium pebbles, fabricated by the rotating electrode method, were neutron irradiated in the HFR in Petten within the HIDOBE-01 experiment. This study presents tritium release and retention properties and data on microstructure evolution of beryllium pebbles irradiated at 630, 740, 873, 948 K up to a damage dose of 18 dpa, corresponding to a helium accumulation of about 3000 appm. The measured cumulative released activity from the beryllium pebbles irradiated at 948 K was found to be significantly lower than the calculated value. After irradiation at 873 and 948 K scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses revealed large pores or bubbles in the bulk and oxide films with a thickness of up to 8 μm at the surface of the beryllium pebbles. The radiation-enhanced diffusion of tritium and the formation of open porosity networks accelerate the tritium release from the beryllium pebbles during the high-flux neutron irradiation.

Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

International Nuclear Information System (INIS)

Aydogdu, K.

1998-01-01

Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

Physical particularities of nuclear reactors using heavy moderators of neutrons

Energy Technology Data Exchange (ETDEWEB)

Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

2016-12-15

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

Physical particularities of nuclear reactors using heavy moderators of neutrons

International Nuclear Information System (INIS)

Kulikov, G. G.; Shmelev, A. N.

2016-01-01

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using "2"3"3U as a fissile nuclide and "2"3"2Th and "2"3"1Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

The influence of the (n, 2n) and (n, {alpha}) reactions of beryllium on the neutron balance in a BeO or Be moderated reactor and its consequences on the long term reactivity changes; Influence des reactions (n, 2n) et (n, {alpha}) du beryllium sur le bilan neutronique d'un reacteur modere a l'oxyde de beryllium ou au beryllium. Consequences sur l'evolution a long terme de la reactivite

Energy Technology Data Exchange (ETDEWEB)

Sahai, K; Benoist, P; Horowitz, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The reaction probabilities in an infinite and homogeneous medium of BeO or Be have been calculated from neutron cross-section curves, for a neutron produced with an energy distribution similar to a fission spectrum; the calculation shows that, after several elastic collisions, the neutron has yet an appreciable probability to undergo a reaction, in spite of the energy degradation in the spectrum due to each collision. This degradation has been calculated, taking into account of anisotropy of the collisions. The gain of the reactivity in a reactor has been obtained after correcting these probabilities for the attenuation of the flux of fission neutrons due to the inelastic scattering in the uranium. Finally, the calculation shows that in a power reactor, this gain of reactivity is in practice destroyed in a few years by the accumulation of poisonous nuclei such as Li{sup 6} and He{sup 3} following (n, {alpha}) reaction. (author) [French] Les probabilites de reaction en milieu infini et homogene de glucine (BeO) ou de beryllium ont ete calculees a partir des courbes de section efficace pour un neutron naissant suivant un spectre de fission; le calcul montre qu'apres plusieurs diffusions elastiques le neutron a encore une probabilite appreciable de subir une reaction, malgre la degradation du spectre a chaque diffusion; cette degradation a ete calculee en tenant compte de l'anisotropie du choc. Le gain de reactivite dans un reacteur a ensuite ete obtenu en corrigeant les probabilites en milieu homogene de l'effet l'attenuation du flux des neutrons de fission par les chocs inelastiques dans les barres d'uranium. Enfin, le calcul montre que, dans un reacteur de puissance, ce gain de reactivite est pratiquement detruit en peu d'annees par l'accumulation de noyaux poisons Li{sup 6} et He{sup 3} consecutive a la reaction (n, {alpha}). (auteur)

Chronic Beryllium Disease

Science.gov (United States)

... who are exposed to beryllium will not experience health effects. Studies have shown that on average, 1 – 6 percent of exposed workers develop beryllium sensitization, although the rates can be ...

Defense programs beryllium good practice guide

Energy Technology Data Exchange (ETDEWEB)

Herr, M.

1997-07-01

Within the DOE, it has recently become apparent that some contractor employees who have worked (or are currently working) with and around beryllium have developed chronic beryllium disease (CBD), an occupational granulomatous lung disorder. Respiratory exposure to aerosolized beryllium, in susceptible individuals, causes an immunological reaction that can result in granulomatous scarring of the lung parenchyma, shortness of breath, cough, fatigue, weight loss, and, ultimately, respiratory failure. Beryllium disease was originally identified in the 1940s, largely in the fluorescent light industry. In 1950, the Atomic Energy Commission (AEC) introduced strict exposure standards that generally curtailed both the acute and chronic forms of the disease. Beginning in 1984, with the identification of a CBD case in a DOE contractor worker, there was increased scrutiny of both industrial hygiene practices and individuals in this workforce. To date, over 100 additional cases of beryllium-specific sensitization and/or CBD have been identified. Thus, a disease previously thought to be largely eliminated by the adoption of permissible exposure standards 45 years ago is still a health risk in certain workforces. This good practice guide forms the basis of an acceptable program for controlling workplace exposure to beryllium. It provides (1) Guidance for minimizing worker exposure to beryllium in Defense Programs facilities during all phases of beryllium-related work, including the decontamination and decommissioning (D&D) of facilities. (2) Recommended controls to be applied to the handling of metallic beryllium and beryllium alloys, beryllium oxide, and other beryllium compounds. (3) Recommendations for medical monitoring and surveillance of workers exposed (or potentially exposed) to beryllium, based on the best current understanding of beryllium disease and medical diagnostic tests available. (4) Site-specific safety procedures for all processes of beryllium that is likely to

Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1984-07-01

It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.

Defense programs beryllium good practice guide

International Nuclear Information System (INIS)

Herr, M.

1997-07-01

Within the DOE, it has recently become apparent that some contractor employees who have worked (or are currently working) with and around beryllium have developed chronic beryllium disease (CBD), an occupational granulomatous lung disorder. Respiratory exposure to aerosolized beryllium, in susceptible individuals, causes an immunological reaction that can result in granulomatous scarring of the lung parenchyma, shortness of breath, cough, fatigue, weight loss, and, ultimately, respiratory failure. Beryllium disease was originally identified in the 1940s, largely in the fluorescent light industry. In 1950, the Atomic Energy Commission (AEC) introduced strict exposure standards that generally curtailed both the acute and chronic forms of the disease. Beginning in 1984, with the identification of a CBD case in a DOE contractor worker, there was increased scrutiny of both industrial hygiene practices and individuals in this workforce. To date, over 100 additional cases of beryllium-specific sensitization and/or CBD have been identified. Thus, a disease previously thought to be largely eliminated by the adoption of permissible exposure standards 45 years ago is still a health risk in certain workforces. This good practice guide forms the basis of an acceptable program for controlling workplace exposure to beryllium. It provides (1) Guidance for minimizing worker exposure to beryllium in Defense Programs facilities during all phases of beryllium-related work, including the decontamination and decommissioning (D ampersand D) of facilities. (2) Recommended controls to be applied to the handling of metallic beryllium and beryllium alloys, beryllium oxide, and other beryllium compounds. (3) Recommendations for medical monitoring and surveillance of workers exposed (or potentially exposed) to beryllium, based on the best current understanding of beryllium disease and medical diagnostic tests available. (4) Site-specific safety procedures for all processes of beryllium that is

Thermomechanical testing of beryllium for the JET/ISX-B beryllium limiter experiment

International Nuclear Information System (INIS)

Watson, R.D.; Smith, M.F.; Whitley, J.B.; McDonald, J.M.

1984-01-01

Materials testing of S-65-B beryllium has been conducted in support of the beryllium limiter experiment on the ISX-B tokamak. The S-65-B grade of hot-pressed beryllium was chosen over S-200-E because of its superior strength and ductility at elevated temperatures. The testing has included measurement of tensile and yield strength, ductility, Young's Modulus, thermal conductivity, and specific heat from 50 0 C to 700 0 C. Thermal fatigue testing of a 2.5 cm beryllium cube was conducted using an electron beam to apply a heat flux of 2.5 kw/cm 2 for 0.3 second pulses for 1500 cycles. Results from the tests are compared to elastic-plastic finite element stress calculations. The testing indicates that the ISX-B beryllium limiter should survive the tokamak environment without serious structural failure, although some surface cracking is expected to occur. (author)

Measurement of the diffusion length of thermal neutrons in the beryllium oxide; Mesure de la longueur de diffusion des neutrons thermiques dans l'oxyde de beryllium

Energy Technology Data Exchange (ETDEWEB)

Koechlin, J C; Martelly, J; Duggal, V P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The diffusion length of thermal neutrons in the beryllium oxide has been obtained while studying the spatial distribution of the neutrons in a massive parallelepiped of this matter placed before the thermal column of the reactor core of Saclay. The mean density of the beryllium oxide (BeO) is 2,95 gr/cm{sup 3}, the mean density of the massif is 2,92 gr/cm{sup 3}. The value of the diffusion length, deducted of the done measures, is: L = 32,7 {+-} 0,5 cm (likely gap). Some remarks are formulated about the influence of the spectral distribution of the neutrons flux used. (authors) [French] La longueur de diffusion des neutrons thermiques dans l'oxyde de beryllium a ete obtenue en etudiant la repartition spatiale des neutrons dans un massif parallelepipedique de cette matiere placee devant la colonne thermique de la Pile de Saclay. La densite moyenne de l'oxyde de beryllium (BeO) est de 2,95 gr/cm{sup 3}, la densite moyenne du massif de 2,92 gr/cm{sup 3}. La valeur de la longueur de diffusion, deduite des mesures effectuees est: L 32,7 {+-} 0,5 cm (ecart probable). Des remarques sont formulees quant a l'influence de la repartition spectrale du flux de neutrons utilise. (auteurs)

Beryllium allergy

International Nuclear Information System (INIS)

Schoenherr, S.; Pevny, I.

1989-12-01

Beryllium is not only a high potent allergen, but also a fotoallergen and can provoke contact allergic reactions, fotoallergic reactions, granulomatous skin reactions, pulmonary granulomatous diseases and sometimes even systemic diseases. The authors present 9 own cases of a patch test positive beryllium allergy, 7 patients with relevant allergy and 5 patients with an allergic contact stomatitis. (author)

Solid state bonding of beryllium-copper for an ITER first wall application

International Nuclear Information System (INIS)

Odegard, B.C. Jr.; Cadden, C.H.

1998-02-01

Several different joint assemblies were evaluated in support of a manufacturing technology for diffusion bonding a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Because beryllium reacts with all but a few elements to form intermetallic compounds, this study considered several different surface treatments as a means of both inhibiting these reactions and promoting a good diffusion bond between the two substrates. A diffusion bonded assemblies used aluminum or an aluminum-beryllium composite (AlBeMet-150) as the interfacial material in contact with beryllium. In most cases, explosive bonding was utilized as a technique for joining the copper alloy heat sink to an aluminum or AlBeMet-150 substrate, which was subsequently diffusion bonded to an aluminum coated beryllium tile. In this approach, a 250 microm thick titanium foil was used as a diffusion barrier between the copper and aluminum to prevent the formation of Cu-Al intermetallic phases. In all cases, a hot isostatic pressing (HIP) furnace was used in conjunction with canned assemblies in order to minimize oxidation and apply sufficient pressure on the assembly for excellent metal-to-metal contact and subsequent bonding. Several different processing schedules were evaluated during the course of this study; bonded assemblies were produced that failed outside the bond area indicating a 100% joint efficiency

Solid state bonding of beryllium-copper for an ITER first wall application

Energy Technology Data Exchange (ETDEWEB)

Odegard, B.C. Jr.; Cadden, C.H. [Sandia National Labs., Livermore, CA (United States)

1998-01-01

Several different joint assemblies were evaluated in support of a manufacturing technology for diffusion bonding a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Because beryllium reacts with all but a few elements to form intermetallic compounds, this study considered several different surface treatments as a means of both inhibiting these reactions and promoting a good diffusion bond between the two substrates. All diffusion bonded assemblies used aluminum or an aluminum-beryllium composite (AlBeMet-150) as the interfacial material in contact with beryllium. In most cases, explosive bonding was utilized as a technique for joining the copper alloy heat sink to an aluminum or AlBeMet-150 substrate, which was subsequently diffusion bonded to an aluminum coated beryllium tile. In this approach, a 250 {mu}m thick titanium foil was used as a diffusion barrier between the copper and aluminum to prevent the formation of Cu-Al intermetallic phases. In all cases, a hot isostatic pressing (HIP) furnace was used in conjunction with canned assemblies in order to minimize oxidation and apply sufficient pressure on the assembly for excellent metal-to-metal contact and subsequent bonding. Several different processing schedules were evaluated during the course of this study; bonded assemblies were produced that failed outside the bond area indicating a 100% joint efficiency. (author)

Heavy water moderated tubular type nuclear reactor

International Nuclear Information System (INIS)

Oohashi, Masahisa.

1986-01-01

Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

Measurement of the diffusion length of thermal neutrons in the beryllium oxide

International Nuclear Information System (INIS)

Koechlin, J.C.; Martelly, J.; Duggal, V.P.

1955-01-01

The diffusion length of thermal neutrons in the beryllium oxide has been obtained while studying the spatial distribution of the neutrons in a massive parallelepiped of this matter placed before the thermal column of the reactor core of Saclay. The mean density of the beryllium oxide (BeO) is 2,95 gr/cm 3 , the mean density of the massif is 2,92 gr/cm 3 . The value of the diffusion length, deducted of the done measures, is: L = 32,7 ± 0,5 cm (likely gap). Some remarks are formulated about the influence of the spectral distribution of the neutrons flux used. (authors) [fr

Conceptual design of a pressure tube light water reactor with variable moderator control

International Nuclear Information System (INIS)

Rachamin, R.; Fridman, E.; Galperin, A.

2012-01-01

This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

Development of uranium dioxide fuel pellets with addition of beryllium oxide for increasing of thermal conductivity

International Nuclear Information System (INIS)

Queiroz, Carolinne Mol; Ferreira, Ricardo Alberto Neto

2011-01-01

The CDTN - Centro de Desenvolvimento de Tecnologia Nuclear presents a project named 'Beryllium Project' viewing to increasing the thermal conductivity of UO 2 fuel pellets, increasing the lifetime of those pellets in the reactor, generating a greater economy. This increase of conductivity is obtained by means of Be O addition to the UO 2 fuel pellets, which is very used for the production of nuclear energy. The UO 2 pellets however present a thermal conductivity relatively low, generating a high temperature gradient between the center and his side surface. The addition of beryllium oxide, with higher thermal conductivity gives pellets which will present lower temperature gradient and, consequently, more durability and better utilization of energy potential of the pellet in the reactor. (author)

Sensitivity analysis of reflector types and impurities in 10 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2007-01-01

The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

Measurements of beryllium sputtering yields at JET

Science.gov (United States)

Jet-Efda Contributors Stamp, M. F.; Krieger, K.; Brezinsek, S.

2011-08-01

The lifetime of the beryllium first wall in ITER will depend on erosion and redeposition processes. The physical sputtering yields for beryllium (both deuterium on beryllium (Be) and Be on Be) are of crucial importance since they drive the erosion process. Literature values of experimental sputtering yields show an order of magnitude variation so predictive modelling of ITER wall lifetimes has large uncertainty. We have reviewed the old beryllium yield experiments on JET and used current beryllium atomic data to produce revised beryllium sputtering yields. These experimental measurements have been compared with a simple physical sputtering model based on TRIM.SP beryllium yield data. Fair agreement is seen for beryllium yields from a clean beryllium limiter. However the yield on a beryllium divertor tile (with C/Be co-deposits) shows poor agreement at low electron temperatures indicating that the effect of the higher sputtering threshold for beryllium carbide is important.

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor

International Nuclear Information System (INIS)

Bretscher, M. M.

1998-01-01

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% 235 U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm 3 and 3.8 gU/cm 3 are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively

Tritium release from beryllium pebbles after high temperature irradiation up to 3000 appm He in the HIDOBE-01 experiment

Energy Technology Data Exchange (ETDEWEB)

Til, S. van, E-mail: vantil@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Fedorov, A.V.; Stijkel, M.P.; Cobussen, H.L.; Mutnuru, R.K.; Idsert, P. van der [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, 1755 ZG Petten (Netherlands); Zmitko, M. [The European Joint Undertaking for ITER and The Development of Fusion Energy, c/ Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, 08019 Barcelona (Spain)

2013-11-15

In the HIDOBE (HIgh DOse irradiation of BEryllium) irradiation program, various grades of constrained and unconstrained beryllium pebbles, beryllium pellets and titanium-beryllide samples are irradiated in the High Flux Reactor (HFR) in Petten at four different temperatures (between 698 K and 1023 K) for 649 days [1]. The first of two HIDOBE irradiation experiments, HIDOBE-01, was completed after achieving a DEMO relevant helium production level of 3000 appm and the samples are retrieved for postirradiation examination (PIE). This work shows preliminary results of the out-of-pile tritium release analysis performed on different grades of irradiated beryllium pebbles (different in size). Relationships between irradiation temperature, tritium inventory and microstructural evolution have been observed by light microscopy and scanning electron microscopy.

Beryllium R&D for blanket application

Science.gov (United States)

Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

1998-10-01

The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

The benchmark experiment on slab beryllium with D–T neutrons for validation of evaluated nuclear data

Energy Technology Data Exchange (ETDEWEB)

Nie, Y., E-mail: nieyb@ciae.ac.cn [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); Ren, J.; Ruan, X.; Bao, J. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); Han, R. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Zhang, S. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Inner Mongolia University for the Nationalities, Inner Mongolia, Tongliao 028000 (China); Huang, H.; Li, X. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); Ding, Y. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000 (China); Wu, H.; Liu, P.; Zhou, Z. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China)

2016-04-15

Highlights: • Evaluated data for beryllium are validated by a high precision benchmark experiment. • Leakage neutron spectra from pure beryllium slab are measured at 61° and 121° using time-of-flight method. • The experimental results are compared with the MCNP-4B calculations with the evaluated data from different libraries. - Abstract: Beryllium is the most favored neutron multiplier candidate for solid breeder blankets of future fusion power reactors. However, beryllium nuclear data are differently presented in modern nuclear data evaluations. In order to validate the evaluated nuclear data on beryllium, in the present study, a benchmark experiment has been performed at China Institution of Atomic Energy (CIAE). Neutron leakage spectra from pure beryllium slab samples were measured at 61° and 121° using time-of-flight method. The experimental results were compared with the calculated ones by MCNP-4B simulation, using the evaluated data of beryllium from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 libraries. From the comparison between the measured and the calculated spectra, it was found that the calculation results based on CENDL-3.1 caused overestimation in the energy range from about 3–12 MeV at 61°, while at 121°, all the libraries led to underestimation below 3 MeV.

The benchmark experiment on slab beryllium with D–T neutrons for validation of evaluated nuclear data

International Nuclear Information System (INIS)

Nie, Y.; Ren, J.; Ruan, X.; Bao, J.; Han, R.; Zhang, S.; Huang, H.; Li, X.; Ding, Y.; Wu, H.; Liu, P.; Zhou, Z.

2016-01-01

Highlights: • Evaluated data for beryllium are validated by a high precision benchmark experiment. • Leakage neutron spectra from pure beryllium slab are measured at 61° and 121° using time-of-flight method. • The experimental results are compared with the MCNP-4B calculations with the evaluated data from different libraries. - Abstract: Beryllium is the most favored neutron multiplier candidate for solid breeder blankets of future fusion power reactors. However, beryllium nuclear data are differently presented in modern nuclear data evaluations. In order to validate the evaluated nuclear data on beryllium, in the present study, a benchmark experiment has been performed at China Institution of Atomic Energy (CIAE). Neutron leakage spectra from pure beryllium slab samples were measured at 61° and 121° using time-of-flight method. The experimental results were compared with the calculated ones by MCNP-4B simulation, using the evaluated data of beryllium from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 libraries. From the comparison between the measured and the calculated spectra, it was found that the calculation results based on CENDL-3.1 caused overestimation in the energy range from about 3–12 MeV at 61°, while at 121°, all the libraries led to underestimation below 3 MeV.

Corrosion of beryllium oxide; Corrosion de l'oxyde de beryllium

Energy Technology Data Exchange (ETDEWEB)

Elston, J; Caillat, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

Data are reported on the volatilization rate of beryllium oxide in moist air depending on temperature and water vapour concentration. They are concerned with powder samples or sintered shapes of various densities. For sintered samples, the volatilization rate is very low under the following conditions: - temperature: 1300 deg. C, - water vapour concentration in moist air: 25 g/m{sup 3}, - flow rate: 12 I/hour corresponding to a speed of 40 m/hour on the surface of the sample. For calcinated powders (1300 deg. C), grain growth has been observed under a stream of moist air at 1100 deg. C. For instance, grain size changes from 0,5 to at least 2 microns after 500 hours of exposure at this temperature. Furthermore, results data are reported on corrosion of sintered beryllium oxide in pressurized water. At 250 deg. C, under a pressure of 40 kg/cm{sup 2} water is very slightly corrosive; however, internal strains are revealed. Finally, some features on the corrosion in liquid sodium are exposed. (author)Fren. [French] La volatilisation de l'oxyde de beryllium dans l'air humide est etudiee en fonction de la temperature pour differentes teneurs de vapeur d'eau. Les essais decrits portent sur de l'oxyde de beryllium en poudre ou sur des echantillons d'oxyde de beryllium fritte de differentes densites. Avec un debit d'air de 12 I/h contenant 25 g de vapeur par m{sup 3} correspondant a une vitesse de 40 m/h sur la surface de l'echantillon, la volatilisation des frittes a 1300 deg. C reste tres faible. Sur de la poudre d'oxyde de beryllium calcinee initialement a 1300 deg. C, on observe un grossissement de la taille des grains sous l'influence de l'air humide a 1100 deg. C. Par exemple, elle passe de 0,5 a au moins 2 microns apres 500 heures d'exposition a cette temperature. On donne d'autre part les resultats d'une etude de la corrosion de frittes d'oxyde de beryllium par l'eau, en autoclave. A 250 deg. C, sous une pression de 40 kg/cm{sup 2}, l'action de l'eau reste tres

Vacuum hot-pressed beryllium and TiC dispersion strengthened tungsten alloy developments for ITER and future fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Liu, Xiang, E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Chen, Jiming; Lian, Youyun; Wu, Jihong; Xu, Zengyu; Zhang, Nianman; Wang, Quanming; Duan, Xuro [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Wang, Zhanhong; Zhong, Jinming [Northwest Rare Metal Material Research Institute, CNMC, Ningxia Orient Group Co. Ltd.,No.119 Yejin Road, Shizuishan City, Ningxia,753000 (China)

2013-11-15

Beryllium and tungsten have been selected as the plasma facing materials of the ITER first wall (FW) and divertor chamber, respectively. China, as a participant in ITER, will share the manufacturing tasks of ITER first-wall mockups with the European Union and Russia. Therefore ITER-grade beryllium has been developed in China and a kind of vacuum hot-pressed (VHP) beryllium, CN-G01, was characterized for both physical, and thermo-mechanical properties and high heat flux performance, which indicated an equivalent performance to U.S. grade S-65C beryllium, a reference grade beryllium of ITER. Consequently CN-G01 beryllium has been accepted as the armor material of ITER-FW blankets. In addition, a modification of tungsten by TiC dispersion strengthening was investigated and a W–TiC alloy with TiC content of 0.1 wt.% has been developed. Both surface hardness and recrystallization measurements indicate its re-crystallization temperature approximately at 1773 K. Deuterium retention and thermal desorption behaviors of pure tungsten and the TiC alloy were also measured by deuterium ion irradiation of 1.7 keV energy to the fluence of 0.5–5 × 10{sup 18} D/cm{sup 2}; a main desorption peak at around 573 K was found and no significant difference was observed between pure tungsten and the tungsten alloy. Further characterization of the tungsten alloy is in progress.

Sensitivity analysis of reflector types and impurities in a 10 MW MTR type nuclear research reactor

International Nuclear Information System (INIS)

Khattab, K.; Khamis, I.

2008-01-01

The 2-D and 3-D neutronics models for 10 MW nuclear research reactor of MTR type have been developed and presented in this paper. Our results agree very well with the results of seven countries mentioned in the IAEA-TECDOC-233. To study the effect of reflector types on the reactor effective multiplication factor, five types of reflectors such as pure beryllium, beryllium, heavy water, carbon and water are selected for this study. The pure beryllium is found to be the most efficient reflector in this group. The effect of the most important impurities, which exist on the beryllium reflector such as iron, silicon and aluminium on the reactor multiplication factor, have been analyzed as well. It is found that the iron impurity affects the reactor multiplication factor the most compared to silicon and aluminium impurities. (author)

Thermal effects on beryllium mirrors

International Nuclear Information System (INIS)

Weinswig, S.

1989-01-01

Beryllium is probably the most frequently used material for spaceborne system scan mirrors. Beryllium's properties include lightweightedness, high Young's modulus, high stiffness value, high resonance value. As an optical surface, beryllium is usually nickel plated in order to produce a higher quality surface. This process leads to the beryllium mirror acting like a bimetallic device. The mirror's deformation due to the bimetallic property can possibly degrade the performance of the associated optical system. As large space borne systems are designed and as temperature considerations become more crucial in the instruments, the concern about temporal deformation of the scan mirrors becomes a prime consideration. Therefore, two sets of tests have been conducted in order to ascertain the thermal effects on nickel plated beryllium mirrors. These tests are categorized. The purpose of this paper is to present the values of the bimetallic effect on typical nickel plated beryllium mirrors

Power control device for heavy water moderated reactor

International Nuclear Information System (INIS)

Matsushima, Hidesuke; Masuda, Hiroyuki.

1978-01-01

Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

Reactivity margins in heavy water moderated production reactors

International Nuclear Information System (INIS)

Benton, F.D.

1981-11-01

The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

Desorption of tritium and helium from high dose neutron irradiated beryllium

Science.gov (United States)

Kupriyanov, I. B.; Nikolaev, G. N.; Vlasov, V. V.; Kovalev, A. M.; Chakin, V. P.

2007-08-01

The effect of high dose neutron irradiation on tritium and helium desorption in beryllium is described. Beryllium samples were irradiated in the SM and BOR-60 reactors to a neutron fluences ( E > 0.1 MeV) of (5-16) × 10 22 cm -2 at 70-100 °C and 380-420 °C. A mass-spectrometry technique was used in out of pile tritium release experiments during stepped annealing in the 250-1300 °C temperature range. The total amount of helium accumulated in irradiated beryllium samples varied from 6000 to 7200 appm. The first signs of tritium and helium release were detected at temperature of 312-445 °C and 500-740 °C, respectively. It is shown that most tritium (˜82%) from sample irradiated at 70-100 °C releases in temperature range of 312-700 °C before the beginning of helium release (740 °C). In the case of beryllium sample irradiated at 380-420 °C, tritium release starts at a higher temperature ( Ts > Tann = 445 °C) and most of the tritium (˜99.8%) is released concurrently with helium which could be considered as evidence of co-existence of partial amounts of tritium and helium in common bubbles. Both the Be samples differ little in the upper temperatures of gas release: 745 and 775 °C for tritium; 1140 and 1160 °C for helium. Swelling of beryllium starts to play a key role in accelerating tritium release at Tann > 600 °C and in helium release - at Tann > 750 °C.

Sintering of beryllium oxide; Frittage de l'oxyde de beryllium

Energy Technology Data Exchange (ETDEWEB)

Caillat, R; Pointud, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

This study had for origin to find a process permitting to manufacture bricks of beryllium oxide of pure nuclear grade, with a density as elevated as possible and with standardized shape. The sintering under load was the technique kept for the manufacture of the bricks. Because of the important toxicity of the beryllium oxide, the general features for the preliminary study of the sintering, have been determined while using alumina. The obtained results will be able to act as general indication for ulterior studies with sintering under load. (M.B.) [French] Cette etude a eu pour origine la recherche d'un procede permettant de fabriquer industriellement des briques d'oxyde de beryllium nucleaireraent pures, de densite aussi elevee que possible et de forme standardisee. Le frittage sous charge fut la technique retenue pour la fabrication des briques. En raison de la grande toxicite de l'oxyde de beryllium, les caracteristiques generales du frittage, pour l'etude preliminaire, ont ete determine en utilisant de l'alumine. Les resultats obtenus pourront servir d'indication generale pour des etudes ulterieurs avec frittage sous charge. (M.B.)

Fracture toughness of irradiated beryllium

International Nuclear Information System (INIS)

Beeston, J.M.

1978-01-01

The fracture toughness of nuclear grade hot-pressed beryllium upon irradiation to fluences of 3.5 to 5.0 x 10 21 n/cm 2 , E greater than 1 MeV, was determined. Procedures and data relating to a round-robin test contributing to a standard ASTM method for unirradiated beryllium are discussed in connection with the testing of irradiated specimens. A porous grade of beryllium was also irradiated and tested, thereby enabling some discrimination between the models for describing the fracture toughness behavior of porous beryllium. The fracture toughness of unirradiated 2 percent BeO nuclear grade beryllium was 12.0 MPa m/sup 1 / 2 /, which was reduced 60 percent upon irradiation at 339 K and testing at 295 K. The fracture toughness of a porous grade of beryllium was 13.1 MPa m/sup 1 / 2 /, which was reduced 68 percent upon irradiation and testing at the same conditions. Reasons for the reduction in fracture toughness upon irradiation are discussed

Neutron absorption profile in a reactor moderated by different mixtures of light and heavy waters

International Nuclear Information System (INIS)

Nagy, Mohamed E.; Aly, Mohamed N.; Gaber, Fatma A.; Dorrah, Mahmoud E.

2014-01-01

Highlights: • We studied neutron absorption spectra in a mixed water moderated reactor. • Changing D 2 O% in moderator induced neutron energy spectral shift. • Most of the neutrons absorbed in control rods were epithermal. • Control rods worth changes were not proportional to changes of D 2 O% in moderator. • Control rod arrangement influenced the neutronic behavior of the reactor. - Abstract: A Monte-Carlo parametric study was carried out to investigate the neutron absorption profile in a model of LR-0 reactor when it is moderated by different mixtures of heavy/light waters at molecular ratios ranging from 0% up to 100% D 2 O at increments of 10% in D 2 O. The tallies included; neutron absorption profiles in control rods and moderator, and neutron capture profile in 238 U. The work focused on neutron absorption in control rods entailing; total mass of control rods needed to attain criticality, neutron absorption density and total neutron absorption in control rods at each of the studied mixed water moderators. The aim was to explore whether thermal neutron poisons are the most suitable poisons to be used in control rods of nuclear reactors moderated by mixed heavy/light water moderators

Power spectral density measurements with 252Cf for a light water moderated research reactor

International Nuclear Information System (INIS)

King, W.T.; Mihalczo, J.T.

1979-01-01

A method of determining the reactivity of far subcritical systems from neutron noise power spectral density measurements with 252 Cf has previously been tested in fast reactor critical assemblies: a mockup of the Fast Flux Test Facility reactor and a uranium metal sphere. Calculations indicated that this measurement was feasible for a pressurized water reactor (PWR). In order to evaluate the ability to perform these measurements with moderated reactors which have long prompt neutron lifetimes, measurements were performed with a small plate-type research reactor whose neutron lifetime (57 microseconds) was about a factor of three longer than that of a PWR and approx. 50% longer than that of a boiling water reactor. The results of the first measurements of power spectral densities with 252 Cf for a water moderated reactor are presented

Fuel enrichment reduction for heavy water moderated research reactors

International Nuclear Information System (INIS)

McCulloch, D.B.

1984-01-01

Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

International Nuclear Information System (INIS)

Pinedo V, J.L.

1979-01-01

One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

International Nuclear Information System (INIS)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

2000-01-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Long-term scenarios of power reactors and fuel cycle development and the role of reduced moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

Reduced moderation spectrum reactor is one of water cooled type reactors in future, which is based on the advanced technology of conventional nuclear power plants. The reduced moderation water reactor (RMWR) has various advantages, such as effective utilization of uranium resources, high conversion ratio, high burn-up, long-term cycle operation, and multiple recycle of plutonium. The RMWR is expected to be a substitute of fast breeder reactor (FBR) of which the development encounters with some technical and financial difficulties, and discontinues in many countries. The role of the RMWR on long-term scenarios of power reactor and fuel cycle development in Japan is investigated from the point of view of uranium resource needed. The consumption of natural uranium needed up to the year 2200 is calculated on various assumptions for the following three cases: (1) no breeder reactor; plutonium-thermal cycle in conventional light water reactor, (2) introduction of the FBR, and (3) introduction of the RMWR. The amounts of natural uranium consumption depends largely on the conversion ratio and plutonium quantity needed of a reactor type. The RMWR has a possibility as a substitute technology of the FBR with the improvement of conversion ratio and high burn-up. (Suetake, M.)

Beryllium-stimulated neopterin as a diagnostic adjunct in chronic beryllium disease.

Science.gov (United States)

Maier, Lisa A; Kittle, Lori A; Mroz, Margaret M; Newman, Lee S

2003-06-01

The diagnosis of chronic beryllium disease (CBD) relies on the beryllium lymphocyte proliferation test (BeLPT) to demonstrate a Be specific immune response. This test has improved early diagnosis, but cannot discriminate beryllium sensitization (BeS) from CBD. We previously found high neopterin levels in CBD patients' serum and questioned whether Be-stimulated neopterin production by peripheral blood cells in vitro might be useful in the diagnosis of CBD. CBD, BeS, Be exposed workers without disease (Be-exp) normal controls and sarcoidosis subjects were enrolled. Peripheral blood mononuclear cells (PBMN) were cultured in the presence and absence of beryllium sulfate. Neopterin levels were determined from cell supernatants by enzyme linked immunosorbent assay (ELISA). Clinical evaluation of CBD subjects included chest radiography, pulmonary function testing, exercise testing, and the BeLPT. CBD patients produced higher levels of neopterin in both unstimulated and Be-stimulated conditions compared to all other subjects (P workplace screening. Copyright 2003 Wiley-Liss, Inc.

OVERVIEW OF BERYLLIUM SAMPLING AND ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

Brisson, M

2009-04-01

Because of its unique properties as a lightweight metal with high tensile strength, beryllium is widely used in applications including cell phones, golf clubs, aerospace, and nuclear weapons. Beryllium is also encountered in industries such as aluminium manufacturing, and in environmental remediation projects. Workplace exposure to beryllium particulates is a growing concern, as exposure to minute quantities of anthropogenic forms of beryllium may lead to sensitization and to chronic beryllium disease, which can be fatal and for which no cure is currently known. Furthermore, there is no known exposure-response relationship with which to establish a 'safe' maximum level of beryllium exposure. As a result, the current trend is toward ever lower occupational exposure limits, which in turn make exposure assessment, both in terms of sampling and analysis, more challenging. The problems are exacerbated by difficulties in sample preparation for refractory forms of beryllium, such as beryllium oxide, and by indications that some beryllium forms may be more toxic than others. This chapter provides an overview of sources and uses of beryllium, health risks, and occupational exposure limits. It also provides a general overview of sampling, analysis, and data evaluation issues that will be explored in greater depth in the remaining chapters. The goal of this book is to provide a comprehensive resource to aid personnel in a wide variety of disciplines in selecting sampling and analysis methods that will facilitate informed decision-making in workplace and environmental settings.

Beryllium coating on Inconel tiles

International Nuclear Information System (INIS)

Bailescu, V.; Burcea, G.; Lungu, C.P.; Mustata, I.; Lungu, A.M.; Rubel, M.; Coad, J.P.; Matthews, G.; Pedrick, L.; Handley, R.

2007-01-01

Full text of publication follows: The Joint European Torus (JET) is a large experimental nuclear fusion device. Its aim is to confine and study the behaviour of plasma in conditions and dimensions approaching those required for a fusion reactor. The plasma is created in the toroidal shaped vacuum vessel of the machine in which it is confined by magnetic fields. In preparation for ITER a new ITER-like Wall (ILW) will be installed on Joint European Torus (JET), a wall not having any carbon facing the plasma [1]. In places Inconel tiles are to be installed, these tiles shall be coated with Beryllium. MEdC represented by the National Institute for Laser, Plasma and Radiation Physics, Magurele, Bucharest and in direct cooperation with Nuclear Fuel Plant Pitesti started to coat Inconel tiles with 8 μm of Beryllium in accordance with the requirements of technical specification and fit for installation in the JET machine. This contribution provides an overview of the principles of manufacturing processes using thermal evaporation method in vacuum and the properties of the prepared coatings. The optimization of the manufacturing process (layer thickness, structure and purity) has been carried out on Inconel substrates (polished and sand blasted) The results of the optimization process and analysis (SEM, TEM, XRD, Auger, RBS, AFM) of the coatings will be presented. Reference [1] Takeshi Hirai, H. Maier, M. Rubel, Ph. Mertens, R. Neu, O. Neubauer, E. Gauthier, J. Likonen, C. Lungu, G. Maddaluno, G. F. Matthews, R. Mitteau, G. Piazza, V. Philipps, B. Riccardi, C. Ruset, I. Uytdenhouwen, R and D on full tungsten divertor and beryllium wall for JET TIER-like Wall Project, 24. Symposium on Fusion Technology - 11-15 September 2006 -Warsaw, Poland. (authors)

Toxicokinetics of beryllium following inhalation of beryllium oxide by Beagle dogs. III

International Nuclear Information System (INIS)

Finch, G.L.; Haley, P.J.; Hoover, M.D.; Mewhinney, J.A.; Bice, D.E.; Eidson, A.F.

1988-01-01

Young adult Beagle dogs inhaled radiolabeled beryllium oxide aerosols ( 7 BeO) prepared at either 500 deg. or 1000 deg. C to achieve one of two initial lung burdens (ILBs) of BeO. After exposure, animals were monitored by whole body counting for 7 Be, and excreta, clinical, and radiographic data were collected. One group of dogs was assigned for serial sacrifice for quantitation of beryllium clearance from lung, translocation to other organs, and histopathologic analysis of lung and lymph nodes. A second group of dogs was subjected to periodic bronchopulmonary lavage for analysis of lymphocyte responsiveness to beryllium. These latter dogs were subsequently re-exposed to the high ILB level of BeO prepared t 500 deg. C. ILBs following the second exposure were higher than that after the first exposure (74 vs. 42 μg BeO/kg, respectively). Except for one dog that exhibited enhanced beryllium retention after the second exposure, patterns of whole body clearance were similar to those observed after the initial exposures to the 500 deg. C-BeO. Analysis of lymphocyte responsiveness to beryllium in the second group of dogs is continuing. (author)

Toxicokinetics of beryllium following inhalation of beryllium oxide by Beagle dogs. III

Energy Technology Data Exchange (ETDEWEB)

Finch, G L; Haley, P J; Hoover, M D; Mewhinney, J A; Bice, D E; Eidson, A F

1988-12-01

Young adult Beagle dogs inhaled radiolabeled beryllium oxide aerosols ({sup 7}BeO) prepared at either 500 deg. or 1000 deg. C to achieve one of two initial lung burdens (ILBs) of BeO. After exposure, animals were monitored by whole body counting for {sup 7}Be, and excreta, clinical, and radiographic data were collected. One group of dogs was assigned for serial sacrifice for quantitation of beryllium clearance from lung, translocation to other organs, and histopathologic analysis of lung and lymph nodes. A second group of dogs was subjected to periodic bronchopulmonary lavage for analysis of lymphocyte responsiveness to beryllium. These latter dogs were subsequently re-exposed to the high ILB level of BeO prepared t 500 deg. C. ILBs following the second exposure were higher than that after the first exposure (74 vs. 42 {mu}g BeO/kg, respectively). Except for one dog that exhibited enhanced beryllium retention after the second exposure, patterns of whole body clearance were similar to those observed after the initial exposures to the 500 deg. C-BeO. Analysis of lymphocyte responsiveness to beryllium in the second group of dogs is continuing. (author)

Reactivity test between beryllium and copper

International Nuclear Information System (INIS)

Kawamura, H.; Kato, M.

1995-01-01

Beryllium has been expected for using as plasma facing material on ITER. And, copper alloy has been proposed as heat sink material behind plasma facing components. Therefore, both materials must be joined. However, the elementary process of reaction between beryllium and copper alloy does not clear in detail. For example, other authors reported that beryllium reacted with copper at high temperature, but it was not obvious about the generation of reaction products and increasing of the reaction layer. In the present work, from this point, for clarifying the elementary process of reaction between beryllium and copper, the out-of-pile compatibility tests were conducted with diffusion couples of beryllium and copper which were inserted in the capsule filled with high purity helium gas (6N). Annealing temperatures were 300, 400, 500, 600 and 700 degrees C, and annealing periods were 100, 300 and 1000h. Beryllium specimens were hot pressed beryllium, and copper specimens were OFC (Oxygen Free Copper)

Method for fabricating beryllium structures

Science.gov (United States)

Hovis, Jr., Victor M.; Northcutt, Jr., Walter G.

1977-01-01

Thin-walled beryllium structures are prepared by plasma spraying a mixture of beryllium powder and about 2500 to 4000 ppm silicon powder onto a suitable substrate, removing the plasma-sprayed body from the substrate and placing it in a sizing die having a coefficient of thermal expansion similar to that of the beryllium, exposing the plasma-sprayed body to a moist atmosphere, outgassing the plasma-sprayed body, and then sintering the plasma-sprayed body in an inert atmosphere to form a dense, low-porosity beryllium structure of the desired thin-wall configuration. The addition of the silicon and the exposure of the plasma-sprayed body to the moist atmosphere greatly facilitate the preparation of the beryllium structure while minimizing the heretofore deleterious problems due to grain growth and grain orientation.

Safe waste management practices in beryllium facilities

International Nuclear Information System (INIS)

Bhat, P.N.; Soundararajan, S.; Sharma, D.N.

2012-01-01

Beryllium, an element with the atomic symbol Be, atomic number 4, has very high stiffness to weight ratio and low density. It has good electrical conductive properties with low coefficient of thermal expansion. These properties make the metal beryllium very useful in varied technological endeavours, However, beryllium is recognised as one of the most toxic metals. Revelation of toxic effects of beryllium resulted in institution of stringent health and safety practices in beryllium handling facilities. The waste generated in such facilities may contain traces of beryllium. Any such waste should be treated as toxic waste and suitable safe waste management practices should be adopted. By instituting appropriate waste management practice and through a meticulously incorporated safety measures and continuous surveillance exercised in such facilities, total safety can be ensured. This paper broadly discusses health hazards posed by beryllium and safe methods of management of beryllium bearing wastes. (author)

Investigation of high purity beryllium for the International Thermonuclear Experimental Reactor (ITER), Task 002. Final report

International Nuclear Information System (INIS)

Vagin, S.P.

1995-05-01

The report includes a description of experimental abilities of Solid Structure Research Laboratory of IAE NNC RK, a results of microstructural characterization of A-4 grade polycrystal Beryllium produced at the Ulba metal plant and a technical project-for irradiation experiments. Technical project contains a detailed description of five proposed experiments, clearing behavior of Beryllium materials under the influence of irradiation, temperature, helium and hydrogen accumulation. Complex irradiation jobs, microstructural investigations and mechanical tests are planned in the framework of these experiments

Beryllium R and D for blanket application

Energy Technology Data Exchange (ETDEWEB)

Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

1998-10-01

The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

Beryllium R and D for blanket application

International Nuclear Information System (INIS)

Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

1998-01-01

The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

Energy Technology Data Exchange (ETDEWEB)

Kurinskiy, P., E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Rolli, R. [Karlsruhe Institute of Technology, Institute for Applied Materials – Materials and Biomechanics (IAM-WBM), P.O. Box 3640, Karlsruhe 76021 (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

2014-10-15

Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA.

The effect of helium generation and irradiation temperature on tritium release from neutron irradiated beryllium

International Nuclear Information System (INIS)

Kupriyanov, I.B.; Gorokhov, V.A.; Vlasov, V.V.; Kovalev, A.M.; Chakin, V.P.

2004-01-01

The effect of neutron irradiation condition on tritium release from beryllium is described in this paper. Beryllium samples were irradiated in the SM reactor with neutron fluence (E > 0.1 MeV) of (0.37-2.0) x 10 22 cm -2 at 70-100degC and 650-700degC. Mass-spectrometer technique was used in out of tritium release experiments during stepped-temperature anneal within a temperature range from 250 to 1300degC. The total amount of helium accumulated in irradiated beryllium samples varied from 521 appm to 3061 appm. The first signs of tritium release were detected at temperature of 406-553degC. It was shown that irradiation temperature and helium generation level significantly affect the tritium release. A fraction of 44 - 74 % of tritium content in samples irradiated at low temperature (70 - 100degC) is release from beryllium at an annealing temperature below 800degC, whereas for samples after high temperature irradiation (650 - 700 degC) tritium release did not exceed 14 %. Majority of tritium (∼68%) is released within a temperature range from 800 to 920 degC. The increase of helium generation from 521 appm to 3061 appm results in lowering the temperature of maximal tritium release rate and the upper temperature of tritium release from beryllium by 100-130degC and 200-240degC, correspondingly. On the basis of data obtained, the diffusion coefficients of tritium in beryllium were calculated. (author)

SOURCE AND PATHWAY DETERMINATION FOR BERYLLIUM FOUND IN BECHTEL NEVADA NORTH LAS VEGAS FACILITIES

Energy Technology Data Exchange (ETDEWEB)

BECHTEL NEVADA

2004-07-01

In response to the report ''Investigation of Beryllium Exposure Cases Discovered at the North Las Vegas Facility of the National Nuclear Security Administration'', published by the U.S. Department of Energy (DOE), National Nuclear Security Administration (NNSA) in August 2003, Bechtel Nevada (BN) President and General Manager Dr. F. A. Tarantino appointed the Beryllium Investigation & Assessment Team (BIAT) to identify both the source and pathway for the beryllium found in the North Las Vegas (NLV) B-Complex. From September 8 to December 18, 2003, the BIAT investigated the pathway for beryllium and determined that a number of locations existed at the Nevada Test Site (NTS) which could have contained sufficient quantities of beryllium to result in contamination if transported. Operations performed in the B-1 Building as a result of characterization activities at the Engine Maintenance, Assembly, and Disassembly (EMAD); Reactor Maintenance, Assembly, and Disassembly (RMAD); Test Cells A and C; and the Central Support Facility in Area 25 had the greatest opportunity for transport of beryllium. Investigative monitoring and sampling was performed at these sites with subsequent transport of sample materials, equipment, and personnel from the NTS to the B-1 Building. The timeline established by the BIAT for potential transport of the beryllium contamination into the B-1 Building was from September 1997 through November 2002. Based on results of recently completed swipe sampling, no evidence of transport of beryllium from test areas has been confirmed. Results less than the DOE beryllium action level of 0.2 ???g/100 cm2 were noted for work support facilities located in Area 25. All of the identified sites in Area 25 worked within the B-1 tenant's residency timeline have been remediated. Legacy contaminants have either been disposed of or capped with clean borrow material. As such, no current opportunity exists for release or spread of beryllium

(Beryllium). Internal Report No. 137, Jan. 15, 1958

International Nuclear Information System (INIS)

Mouret, P.; Rigaud, A.

1959-01-01

After a brief summary of the physical and chemical properties of beryllium, the various chemical treatments which can be applied to beryllium minerals either directly or after a physical enrichment are discussed. These various treatments give either the hydroxide or beryllium salts, from which either beryllium oxide or metallic beryllium can easily be obtained. The purification, analysis and uses of beryllium are also briefly discussed. (author)

Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

International Nuclear Information System (INIS)

King, Jeffrey C.; Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F.

2015-01-01

The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW e and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k eff = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

Moderator configuration options for a low-enriched uranium fueled Kilowatt-class Space Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

King, Jeffrey C., E-mail: kingjc@mines.edu [Nuclear Science and Engineering Program, Colorado School of Mines (CSM), Golden, CO (United States); Mencarini, Leonardo de Holanda; Guimaraes, Lamartine N. F., E-mail: guimaraes@ieav.cta.br, E-mail: mencarini@ieav.cta.br [Instituto de Estudos Avancados (IEAV), Sao Jose dos Campos, SP (Brazil). Divisao de Energia Nuclear

2015-07-01

The Brazilian Air Force, through its Institute for Advanced Studies (Instituto de Estudos Avancados, IEAv/DCTA), and the Colorado School of Mines (CSM) are studying the feasibility of a space nuclear reactor with a power of 1-5 kW{sub e} and fueled with Low-Enriched Uranium (LEU). This type of nuclear reactor would be attractive to signatory countries of the Non-Proliferation Treaty (NPT) or commercial interests. A LEU-fueled space reactor would avoid the security concerns inherent with Highly Enriched Uranium (HEU) fuel. As an initial step, the HEU-fueled Kilowatt Reactor Using Stirling Technology (KRUSTY) designed by the Los Alamos National Laboratory serves as a basis for a similar reactor fueled with LEU fuel. Using the computational code MCNP6 to predict the reactor neutronics performance, the size of the resulting reactor fueled with 19.75 wt% enriched uranium-10 wt% molybdenum alloy fuel is adjusted to match the excess reactivity of KRUSTY. Then, zirconium hydride moderator is added to the core to reduce the size of the reactor. This work presents the preliminary results of the computational modeling, with special emphasis on the comparison between homogeneous and heterogeneous moderator systems, in terms of the core diameter required to meet a specific multiplication factor (k{sub eff} = 1.035). This comparison illustrates the impact of moderator configuration on the size and performance of a LEU-fueled kilowatt-class space nuclear reactor. (author)

Study beryllium microplastic deformation

International Nuclear Information System (INIS)

Papirov, I.I.; Ivantsov, V.I.; Nikolaenko, A.A.; Shokurov, V.S.; Tuzov, Yu.V.

2015-01-01

Microplastic flow characteristics systematically studied for different varieties beryllium. In isostatically pressed beryllium it decreased with increasing particle size of the powder, increasing temperature and increasing the pressing metal purity. High initial values of the limit microelasticity and microflow in some cases are due a high level of internal stresses of thermal origin and over time it can relax slowly. During long-term storage of beryllium materials with high initial resistance values microplastic deformation microflow limit and microflow stress markedly reduced, due mainly to the relaxation of thermal microstrain

Preparation of beryllium hydride

International Nuclear Information System (INIS)

Lowrance, B.R.

1975-01-01

A process is described for the preparation of beryllium hydride which comprises pyrolyzing, while in solution in a solvent inert under the reaction conditions, with respect to reactants and products and at a temperature in the range of about 100 0 to about 200 0 C, sufficient to result in the formation of beryllium hydride, a di-t-alkyl beryllium etherate wherein each tertiary alkyl radical contains from 4 to 20 carbon atoms. The pyrolysis is carried out under an atmosphere inert under the reaction conditions, with respect to reactants and products. (U.S.)

Parametric study of moderator heat exchanger for Candu 6 advanced reactor

International Nuclear Information System (INIS)

Umar, Efrizon; Vecchiarelli, Jack

2000-01-01

The passive moderator system for Candu 6 advanced reactor require moderator heat exchanger with the small size and the low resistance coefficient of the shell-side. The study is to determine the required size of moderator heat exchanger, and to calculate the shell side of resistance coefficient have been done. Using computer code CATHENA, it is concluded that the moderator heat exchanger can be used at full power-normal operation condition, especially for the cases with 3600 to 8100 number of tube and 15.90 mm tube diameter. This study show that the proposed moderator heat exchanger have given satisfactory results

Current status of nuclear research reactor management and utilization program in Thailand

International Nuclear Information System (INIS)

Aramrattana, M.; Busamongkol, Y.

1999-01-01

The TRR1/M1 is the first research reactor and has been in operational for more than 20 years. During the three decades of research reactor operation in Thailand the utilization of research reactor have been broadened in different fields such as agriculture, medicine and industry. Limitation on utilization of the existing reactor in various fields has led to establishing of a new nuclear research center, Ongkharak Nuclear Research Center (ONRC). The ONRC comprises three major facilities, namely Reactor Island, Isotope Production Facility and Waste Processing and Storage Facility. The reactor itself is a 10 MW TRIGA-type fuels, moderated and cooled by light water with beryllium and heavy water as the reflectors. It is a multi-purpose reactor consisting of different facilities inside and around the core for radioisotope production, medical and industrial uses; and for beam experiments such as High Resolution Powder Diffractometry (HRPD), Neutron Radiography (NR), Prompt Gamma Neutron Activation Analysis (PGNAA), and Boron Neutron Capture Therapy (BNCT). The center is expected to be operational by year 2001. (author)

Preparation of beryllium hydride

International Nuclear Information System (INIS)

Roberts, C.B.

1975-01-01

A process is described for preparing beryllium hydride by the direct reaction of beryllium borohydride and aluminum hydride trimethylamine adduct. Volatile by-products and unreacted reactants are readily removed from the product mass by sublimation and/or evaporation. (U.S.)

Conceptual designing of reduced-moderation water reactor with heavy water coolant

Energy Technology Data Exchange (ETDEWEB)

Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

2001-12-01

The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

International Nuclear Information System (INIS)

Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

1995-01-01

This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs

The INEL beryllium multiplication experiment

International Nuclear Information System (INIS)

Smith, J.R.; King, J.J.

1991-03-01

The experiment to measure the multiplication of 14-MeV neutrons in bulk beryllium has been completed. The experiment consists of determining the ratio of 56 Mn activities induced in a large manganese bath by a central 14-MeV neutron source, with and without a beryllium sample surrounding the source. In the manganese bath method a neutron source is placed at the center of a totally-absorbing aqueous solution of MnSo 4 . The capture of neutrons by Mn produces a 56 Mn activity proportional to the emission rate of the source. As applied to the measurement of the multiplication of 14- MeV neutrons in bulk beryllium, the neutron source is a tritium target placed at the end of the drift tube of a small deuteron accelerator. Surrounding the source is a sample chamber. When the sample chamber is empty, the neutrons go directly to the surrounding MnSO 4 solution, and produce a 56 Mn activity proportional to the neutron emission rate. When the chamber contains a beryllium sample, the neutrons first enter the beryllium and multiply through the (n,2n) process. Neutrons escaping from the beryllium enter the bath and produce a 56 Mn activity proportional to the neutron emission rate multiplied by the effective value of the multiplication in bulk beryllium. The ratio of the activities with and without the sample present is proportional to the multiplication value. Detailed calculations of the multiplication and all the systematic effects were made with the Monte Carlo program MCNP, utilizing both the Young and Stewart and the ENDF/B-VI evaluations for beryllium. Both data sets produce multiplication values that are in excellent agreement with the measurements for both raw and corrected values of the multiplication. We conclude that there is not real discrepancy between experimental and calculated values for the multiplication of neutrons in bulk beryllium. 12 figs., 11 tabs., 18 refs

Beryllium minerals - demand strong for miniaturisation

International Nuclear Information System (INIS)

Griffiths, J.

1985-01-01

Beryllium is an essential constituent of over 40 minerals of which two are exploited commercially. Beryl is largely produced in the USSR and China and bertrandite in the U.S.A. Phenacite, from Canada, is also under investigation. The largest extraction plant for the recovery of beryllium in the western world is in Utah, U.S.A. and the company also produces beryllium oxide used in the manufacture of ceramics widely used in the electronics industry and for refractory articles. Beryllium-copper alloys in strip, rod and tube form are produced in the U.S.A., Germany and the U.K. Beryllium ceramics are important because of their high thermal conductivity, electrical insulation, strength and rigidity. The alloys, used as electric connectors, microswitch contacts are important for their high suitability for miniaturisation. The future growth potential for the beryllium industry is in the automotive industries in Europe and Japan. (U.K.)

UK methods for studying fuel management in water moderated reactors

International Nuclear Information System (INIS)

Fayers, F.J.

1970-10-01

Current UK methods for studying fuel management and for predicting the reactor physics performance for both light and heavy water moderated power reactors are reviewed. Brief descriptions are given of the less costly computer codes used for initial assessment studies, and also the more elaborate programs associated with detailed evaluation are discussed. Some of the considerations influencing the accuracy of predictions are included with examples of various types of experimental confirmation. (author)

Comprehensive Measurement of Neutron Yield Produced by 62 MeV Protons on Beryllium Target

International Nuclear Information System (INIS)

Osipenko, M.; Ripani, M.; Ricco, G.; Alba, R.; Schillaci, M.; Cosentino, L.; Del Zoppo, A.; Di Pietro, A.; Figuera, P.; Finocchiaro, P.; Maiolino, C.; Santonocito, D.; Scuderi, V.; Barbagallo, M.; Colonna, N.; Boccaccio, P.; Esposito, J.; Celentano, A.; Viberti, C.M.; Kostyukov, A.

2013-06-01

A low-power prototype of neutron amplifier, based on a 70 MeV, high current proton cyclotron being installed at LNL for the SPES RIB facility, was recently proposed within INFN-E project. This prototype uses a thick Beryllium converter to produce a fast neutron spectrum feeding a sub-critical reactor core. To complete the design of such facility the new measurement of neutron yield from a thick Beryllium target was performed at LNS. This measurement used liquid scintillator detectors to identify produced neutrons by Pulse Shape Discrimination and Time of Flight technique to measure neutron energy in the range 0.5-62 MeV. To extend the covered neutron energy range 3 He detector was used to measure neutrons below 0.5 MeV. The obtained yields were normalized to the charge deposited by the proton beam on the metallic Beryllium target. These techniques allowed to achieve a wide angular coverage from 0 to 150 degrees and to explore almost complete neutron energy interval. (authors)

Calculation of fuel and moderator temperature coefficients in APR1400 nuclear reactor by MVP code

International Nuclear Information System (INIS)

Pham Tuan Nam; Le Thi Thu; Nguyen Huu Tiep; Tran Viet Phu

2014-01-01

In this project, these fuel and moderator temperature coefficients were calculated in APR1400 nuclear reactor by MVP code. APR1400 is an advanced water pressurized reactor, that was researched and developed by Korea Experts, its electric power is 1400 MW. The neutronics calculations of full core is very important to analysis and assess a reactor. Results of these calculation is input data for thermal-hydraulics calculations, such as fuel and moderator temperature coefficients. These factors describe the self-safety characteristics of nuclear reactor. After obtaining these reactivity parameters, they were used to re-run the thermal hydraulics calculations in LOCA and RIA accidents. These thermal-hydraulics results were used to analysis effects of reactor physics parameters to thermal hydraulics situation in nuclear reactors. (author)

Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; Longhurst, G.R. [Lockheed Martin Idaho Technol. Co., Idaho Falls, ID (United States). Idaho Nat. Eng. and Environ. Lab.; Causey, R.A.; Wampler, W.R.; Wilson, K.L. [Sandia National Laboratories, Livermore, CA (United States)]|[Sandia National Labs., Albuquerque, NM (United States); Davis, J.W.; Haasz, A.A. [Institute for Aerospace Studies, University of Toronto, Toronto (Canada); Doerner, R.P. [California Univ., San Diego, La Jolla, CA (United States). Center for Magnetic Recording Research; Federici, G. [ITER JWS Garching Co-center, Garching (Germany)

1999-06-01

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions, show a difference of several orders of magnitude in both inventory and permeation rate to the coolant. (orig.) 78 refs.

Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

International Nuclear Information System (INIS)

Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C.

2017-01-01

The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO_2 fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D_2O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α"M_T(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

International Nuclear Information System (INIS)

Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

2013-01-01

Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively

Study on core design for reduced-moderation water reactors

International Nuclear Information System (INIS)

Okubo, Tsutomu

2002-01-01

The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

Study on core design for reduced-moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

Beryllium. Its minerals. Pt. 1

International Nuclear Information System (INIS)

Lires, O.A.; Delfino, C.A.; Botbol, J.

1990-01-01

With this work a series of reports begins, under the generic name 'Beryllium', related to several aspects of beryllium technology. The target is to update, with critical sense, current bibliographic material in order to be used in further applications. Some of the most important beryllium ores, the Argentine emplacement of their deposits and world occurrence are described. Argentine and world production, resources and reserves are indicated here as well. (Author) [es

Concept of safe tank-type water cooled and moderated reactor with HTGR microparticle fuel compacts

International Nuclear Information System (INIS)

Gol'tsev, A.O.; Kukharkin, N.E.; Mosevitskij, I.S.; Ponomarev-Stepnoj, N.N.; Popov, S.V.; Udyanskij, Yu.N.; Tsibul'skij, V.F.

1993-01-01

Concept of safe tank-type water-cooled and moderated reactor on the basis of HTGR fuel microparticles which enable to avoid environment contamination with radioactive products under severe accidents, is proposed. Results of neutron-physical and thermal-physical studies of water cooled and moderated reactor with HTGR microparticle compacts are presented. Characteristics of two reactors with thermal power of 500 and 1500 MW are indicated within the concept frames. The reactor behaviour under severe accident connected with complete loss of water coolant is considered. It is shown that under such an accident the fission products release from fuel microparticles does not occur

Nuclear calculation methods for light water moderated reactors

International Nuclear Information System (INIS)

Hicks, D.

1961-02-01

This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)

Method for hot pressing beryllium oxide articles

Science.gov (United States)

Ballard, Ambrose H.; Godfrey, Jr., Thomas G.; Mowery, Erb H.

1988-01-01

The hot pressing of beryllium oxide powder into high density compacts with little or no density gradients is achieved by employing a homogeneous blend of beryllium oxide powder with a lithium oxide sintering agent. The lithium oxide sintering agent is uniformly dispersed throughout the beryllium oxide powder by mixing lithium hydroxide in an aqueous solution with beryllium oxide powder. The lithium hydroxide is converted in situ to lithium carbonate by contacting or flooding the beryllium oxide-lithium hydroxide blend with a stream of carbon dioxide. The lithium carbonate is converted to lithium oxide while remaining fixed to the beryllium oxide particles during the hot pressing step to assure uniform density throughout the compact.

Belgian research on fusion beryllium waste

International Nuclear Information System (INIS)

Druyts, F.; Mallants, D.; Sillen, X.; Iseghem, P. Van

2004-01-01

Future fusion power plants will generate important quantities of neutron irradiated beryllium. Although recycling is the preferred management option for this waste, this may not be technically feasible for all of the beryllium, because of its radiological characteristics. Therefore, at SCK·CEN, we initiated a research programme aimed at studying aspects of the disposal of fusion beryllium, including waste characterisation, waste acceptance criteria, conditioning methods, and performance assessment. One of the main issues to be resolved is the development of fusion-specific waste acceptance criteria for surface or deep geological disposal, in particular with regard to the tritium content. In case disposal is the only solution, critical nuclides can be immobilised by conditioning the waste. As a first approach to immobilising beryllium waste, we investigated the vitrification of beryllium. Corrosion tests were performed on both metallic and vitrified beryllium to provide source data for performance assessment. Finally, a first step in performance assessment was undertaken. (author)

Potential exposures and risks from beryllium-containing products.

Science.gov (United States)

Willis, Henry H; Florig, H Keith

2002-10-01

Beryllium is the strongest of the lightweight metals. Used primarily in military applications prior to the end of the Cold War, beryllium is finding new applications in many commercial products, including computers, telecommunication equipment, and consumer and automotive electronics. The use of beryllium in nondefense consumer applications is of concern because beryllium is toxic. Inhalation of beryllium dust or vapor causes a chronic lung disease in some individuals at concentrations as low as 0.01 microg/m3 in air. As beryllium enters wider commerce, it is prudent to ask what risks this might present to the general public and to workers downstream of the beryllium materials industry. We address this question by evaluating the potential for beryllium exposure from the manufacturing, use, recycle, and disposal of beryllium-containing products. Combining a market study with a qualitative exposure analysis, we determine which beryllium applications and life cycle phases have the largest exposure potential. Our analysis suggests that use and maintenance of the most common types of beryllium-containing products do not result in any obvious exposures of concern, and that maintenance activities result in greater exposures than product use. Product disposal has potential to present significant individual risks, but uncertainties concerning current and future routes of product disposal make it difficult to be definitive. Overall, additional exposure and dose-response data are needed to evaluate both the health significance of many exposure scenarios, and the adequacy of existing regulations to protect workers and the public. Although public exposures to beryllium and public awareness and concern regarding beryllium risks are currently low, beryllium risks have psychometric qualities that may lead to rapidly heightened public concern.

Neutron energy spectrum flux profile of Ghana's miniature neutron source reactor core

International Nuclear Information System (INIS)

Sogbadji, R.B.M.; Abrefah, R.G.; Ampomah-Amoako, E.; Agbemava, S.E.; Nyarko, B.J.B.

2011-01-01

Highlights: → The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was studied. → Using 20,484 energy grids, the thermal, slowing down and fast neutron energy regions were studied. - Abstract: The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) x 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) x 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) x 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939-3.7880) x 10 -08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) x 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 x 10 -01 MeV and 8.2680 x 10 -01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) x 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

Materials science problems of blankets in Russian concept of fusion reactor

International Nuclear Information System (INIS)

Solonin, M.I.

1998-01-01

Structural materials, beryllium and tritium breeding materials proposed for blanket of Russian reactor DEMO and Test Modules for ITER are discussed. Main requirements for the materials are concerned with basis current designs of blankets and modules and possibility meet of ones for presence and developed alloys and materials discussed considered. Main properties and results of test of ferrite-martensite and vanadium alloys for DEMO and Test Modules are cited. Beryllium compositions used as component of first wall and neutron multiplier are discussed. Liquid lithium and ceramic (lithium orthosilicate) are treated as tritium breeding materials. Russian development of reactor experimental unit for tritium breeding zone using beryllium, lithium ceramic and ferrite-martensite alloys for structural materials is presented. (orig.)

Measurement of neutron yield by 62 MeV proton beam on a thick Beryllium target

International Nuclear Information System (INIS)

Alba, R; Cosentino, G; Zoppo, A Del; Pietro, A Di; Figuera, P; Finocchiaro, P; Maiolino, C; Santonocito, D; Schillaci, M; Barbagallo, M; Colonna, N; Boccaccio, P; Esposito, J; Celentano, A; Osipenko, M; Ricco, G; Ripani, M; Viberti, C M; Kostyukov, A

2013-01-01

In the framework of research on IVth generation reactors and high intensity neutron sources a low-power prototype neutron amplifier was recently proposed by INFN. It is based on a low-energy, high current proton cyclotron, whose beam, impinging on a thick Beryllium converter, produces a fast neutron spectrum. The world database on the neutron yield from thick Beryllium target in the 70 MeV proton energy domain is rather scarce. The new measurement was performed at LNS, covering a wide angular range from 0 to 150 degrees and an almost complete neutron energy interval. In this contribution the preliminary data are discussed together with the proposed ADS facility.

Research Reactors Types and Utilization

International Nuclear Information System (INIS)

Abdelrazek, I.D.

2008-01-01

A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.

METHOD OF BRAZING BERYLLIUM

Science.gov (United States)

Hanks, G.S.; Keil, R.W.

1963-05-21

A process is described for brazing beryllium metal parts by coating the beryllium with silver (65- 75 wt%)-aluminum alloy using a lithium fluoride (50 wt%)-lithium chloride flux, and heating the coated joint to a temperature of about 700 un. Concent 85% C for about 10 minutes. (AEC)

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

International Nuclear Information System (INIS)

Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han

2013-01-01

The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.

Sanitary-hygienic and ecological aspects of beryllium production

Energy Technology Data Exchange (ETDEWEB)

Dvinskykh, E.M.; Savchuk, V.V.; Sidorov, V.L.; Slobodin, D.B.; Tuzov, Y.V. [Ulba Metallurgical Plant, Ust-Kamenogorsk (Kazakhstan)

1998-01-01

The Report describes An organization of sanitary-hygienic and ecological control of beryllium production at Ulba metallurgical plant. It involves: (1) the consideration of main methods for protection of beryllium production personnel from unhealthy effect of beryllium, (2) main kinds of filters, used in gas purification systems at different process areas, (3) data on beryllium monitoring in water, soil, on equipment. This Report also outlines problems connected with designing devices for a rapid analysis of beryllium in air as well as problems of beryllium production on ecological situation in the town. (author)

Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

Energy Technology Data Exchange (ETDEWEB)

Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-07-01

The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

Design and safety considerations for the 10 MW(t) multipurpose TRIGA reactor in Thailand

International Nuclear Information System (INIS)

Razvi, J.; Bolin, J.M.; Saurwein, J.J.; Whittemore, W.L.; Proongmuang, S.

1999-01-01

General Atomics (GA) is constructing the Ongkharak Nuclear Research Center (ONRC) near Bangkok, Thailand for the Office of Atomic Energy for Peace. The ONRC complex includes the following: A multipurpose 10 MW(t) research reactor; An Isotope Production Facility; Centralized Radioactive Waste Processing and Storage Facilities. The Center is being built 60-km northeast of Bangkok, with a 10 MW(t) TRIGA type research reactor as the centerpiece. Facilities are included for neutron transmutation doping of silicon, neutron capture therapy neutron beam research and for production of a variety of radioisotopes. The facility will also be utilized for applied research and technology development as well as training in reactor operations, conduct of experiments and in reactor physics. The multipurpose, pool-type reactor will be fueled with high-density (45 wt%), low-enriched (19.7 wt%) uranium-erbium-zirconium-hydride (UErZrH) fuel rods, cooled and moderated by light water, and reflected by beryllium and heavy water. The general arrangement of the reactor and auxiliary pool structure allows irradiated targets to be transferred entirely under water from their irradiation locations to the hot cell, then pneumatically transferred to the adjacent Isotope Production Facility for processing. The core configuration includes 4 x 4 array standard TRIGA fuel clusters, modified clusters to serve as fast-neutron irradiation facilities, control rods and an in-core Ir-192 production facility. The active core is reflected on two sides by beryllium and on the other two sides by D 2 O. Additional irradiation facilities are also located in the beryllium reflector blocks and the D 2 O reflector blanket. The fuel provides the fundamental safety feature of the ONRC reactor, and as a result of all the well established accident-mitigating characteristics of the UErZrH fuel itself (large prompt negative temperature coefficient of reactivity, fission product retention and chemical stability), a

The nature of beryllium disease

International Nuclear Information System (INIS)

Williams, W.J.

1977-01-01

The increasing use of beryllium in modern industry poses a continuing health hazard with a real risk of producing incapacitating disease and even death. Beryllium and its salts are very toxic, even in small doses and may produce lesions in any organ. The majority of cases follow inhalation and may cause either acute or chronic lung disease. Acute pulmonary disease is a form of chemical pneumonitis while the chronic disease is characterised by the production of granulomas and fibrosis. The skin may be affected with the finding of dermatitis, acute or chronic ulceration. Other organs commonly involved include the liver and kidneys. The pathology of beryllium disease is not specific and diagnosis depends on satisfying the following criteria - history of exposure, consistent clinical, radiographic and pathological finding, presence of beryllium in tissue/fluid and evidence of hypersensitivity. Recent development of 'in vitro' tests of hypersensitivity may prove of value in both diagnosis and prevention of disease. Beryllium disease responds to steroid therapy but the only sure treatment is avoidance of exposure. (author)

Status of the European R and D on beryllium as multiplier material for breeder blankets

International Nuclear Information System (INIS)

Moeslang, A.; Boccaccini, L.V.; Rabaglino, E.; Piazza, G.; Cardella, A.; Sannen, L.; Scibetta, M.; Laan, J. van der; Hegeman, J.B.J.W.

2004-01-01

Within the international fusion community a variety of breeding blanket concepts are being considered, ranging from more conservative concepts to higher-risk concepts for fusion power reactors. In Europe, the Helium Cooled Pebble Bed (HCPB) blanket is one of the two reference concepts which will also be tested as Test Blanket Module (TBM) in ITER. In addition to the R and D for structural parts of the HCPB blanket, a considerable effort is devoted to the production and qualification of ceramic breeder and neutron multiplier (beryllium or beryllide) pebble beds. Since in the HCPB blanket pebbles made of lithium ceramics are foreseen, a high volume fraction of beryllium as a neutron multiplier to Li-based ceramic of about 4: l is needed. The typical loading conditions for beryllium are, with a neutron wall load of ∼12.5 MWa/m 2 and in ∼5 years lifetime: T min ∼300degC, T max ∼600-900degC, displacement damage ∼80 dpa, peak 4 He production ∼26000 appm and peak 3 H production ∼700 appm at the End-Of-Life. The behaviour of beryllium under irradiation is considered to be a key issue of the HCPB blanket, because of swelling due to helium bubbles and tritium retention. A large R and D programme on beryllium has been implemented in Europe, aimed at characterising and predicting the material behaviour before and under irradiation. An overview on experimental and modelling activities performed during the past 2 years is given with typical results on non-irradiated and irradiated Beryllium materials and pebble beds and the relevance of major results on future beryllium R and D is addressed. (author)

Multi-purpose reactor

International Nuclear Information System (INIS)

1991-05-01

The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

Beryllium electrodeposition on aluminium cathode from chloride melts

International Nuclear Information System (INIS)

Nichkov, I.F.; Novikov, E.A.; Serebryakov, G.A.; Kanashin, Yu.P.; Sardyko, G.N.

1980-01-01

Cathodic processes during beryllium deposition on liquid and solid aluminium cathodes are investigated. Mixture of sodium, potassium and beryllium chloride melts served as an lectrolyte. Beryllium ion discharge at the expense of alloy formation takes place at more positive potentials than on an indifferent cathode at low current densities ( in the case of liquid aluminium cathode). Metallographic analysis and measurements of microhardness have shown, that the cathodic product includes two phases: beryllium solid solution in aluminium and metallic beryllium. It is concluded, that aluminium-beryllium alloys with high cathodic yield by current can be obtained by the electrolytic method

Beryllium satellite thrust cone design, manufacture and test

International Nuclear Information System (INIS)

Schneiter, H.; Chandler, D.

1977-01-01

Pre-formed beryllium sheet material has been used in the design, manufacturing and test of a satellite thrust cone structure. Adhesive bonding was used for attachment of aluminium flanges and conical segment lap strips. Difficulties in beryllium structure design such as incompatibilities with aluminium and handling problems are discussed. Testing to optimize beryllium-beryllium and beryllium-aluminium adhesive bonds is described. The completed thrust cone assembly has been subjected to static load testing and the results are presented. A summary of the relative merits of the use of beryllium in satellite structures is given with recommendations for future users. (author)

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

Walter, C.E.; Pearson, J.S.

1987-05-01

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

Control of beryllium powder at a DOE facility

International Nuclear Information System (INIS)

Langner, G.C.; Creek, K.L.; Castro, R.G.

1997-01-01

Beryllium is contained in a number of domestic and national defense items. Although many items might contain beryllium in some manner, few people need worry about the adverse effects caused by exposure to beryllium because it is the inhalable form of beryllium that is most toxic. Chronic beryllium disease (CBD), a granulomas and fibrotic lung disease with long latency, can be developed after inhalation exposures to beryllium. It is a progressive, debilitating lung disease. Its occurrence in those exposed to beryllium has been difficult to predict because some people seem to react to low concentration exposures whereas others do not react to high concentration exposures. Onset of the disease frequently occurs between 15 to 20 years after exposure begins. Some people develop the disease after many years of low concentration exposures but others do not develop CBD even though beryllium is shown to be present in lungs and urine. Conclusions based on these experiences are that their is some immunological dependence of developing CBD in about 3--4% of the exposed population, but the exact mechanism involved has not yet been identified. Acute beryllium disease can occur after a single exposure to a concentration of greater than 0.100 mg/m3 (inhalation exposure); it is characterized by the development of chemical pneumoconiosis, a respiratory disease. The acute effect of skin contact is a dermatitis characterized by itching and reddened, elevated, or fluid-accumulated lesions which appear particularly on the exposed surfaces of the body, especially the face, neck, arms, and hands. Small particles of beryllium that enter breaks in the skin can lead to the development of granulomas and/or open sores that do not heal until the beryllium has been removed. Our interest is only airborne beryllium, which is found in areas that machine or produce beryllium

Experimental studies of some of the physical features of beryllium-moderated intermediate reactors; Etude experimentale de quelques particularites physiques des reacteurs a neutrons intermediaires, ralentis au beryllium; Ehksperimental'ny e issledovaniya nekotorykh fizicheskikh osobennostej promezhutochnykh reaktorov s berillievym zamedlitelem; Estudios experimentales de algunas caracteristicas fisicas de los reactores intermedios moderados con berilio

Energy Technology Data Exchange (ETDEWEB)

Lejpunskij, A I; Kuznetsov, V A; Artyukhov, G Ya; Mogil' ner, A I; Prokhorov, Yu A; Steklovski, V M; Chernov, L A [Akademiya Nauk, Moskva, Union of Soviet Socialist Republics (Russian Federation)

1962-03-15

This paper is devoted to a review of the results obtained from a number of experiments carried out on the PF-4 critical assembly (intermediate-physical assembly), which is designed for detailed studies of the physical characteristics of intermediate reactors. The cores and reflectors of the various critical assemblies were comprised of compact units of steel or aluminium tubes, in which discs of various materials were placed. By combining 90%-enriched uranium discs with moderating materials in various proportions, and also by introducing moderating layers of different thicknesses into the reflector, it was possible to alter the spectrum of the fission-inducing neutrons within a very broad range. This paper describes the PF-4 critical assembly and its various subassemblies. The relative efficiency of internal and external moderation is analysed for reactors with a very low ratio of moderator nuclei to uranium nuclei in the core. The experiments show that even when thick moderating reflectors are used, this low ratio (the ratio of beryllium nuclei to uranium-235 nuclei being {partial_derivative}Be/{partial_derivative}U{sup 235}{approx_equal}1) leads to an increase of the critical mass. A considerable part of the paper is devoted to an analysis of heterogeneous effects in intermediate reactors. It is shown that for reactors with {partial_derivative}Be/{partial_derivative}U{sup 235}=30-40 various thicknesses of highly enriched uranium, ranging from 0.023 g/cm{sup 2} to 32 g/cm{sup 2}, have an identical effect on the reactivity of the system. The causes underlying compensation of the neutron-flux screening effect by thick layers of uranium are analysed. The interesting fact that the efficiency of uranium increases in the neighbourhood of the absorbing rods, which was experimentally revealed in an assembly with {partial_derivative}Be/{partial_derivative}U{sup 235}{approx_equal}200, is discussed in the paper. This fact is explained by the sharp decline in the importance

Interatomic potentials for fusion reactor material simulations

International Nuclear Information System (INIS)

Bjoerkas, C.

2009-01-01

In this thesis, the behaviour of a material situated in a fusion reactor was studied using molecular dynamics simulations. Simulations of processes in the next generation fusion reactor ITER include the reactor materials beryllium, carbon and tungsten as well as the plasma hydrogen isotopes. This means that interaction models, i.e. interatomic potentials, for this complicated quaternary system are needed. The task of finding such potentials is nonetheless nearly at its end, since models for the beryllium-carbon-hydrogen interactions were constructed in this thesis and as a continuation of that work, a beryllium-tungsten model is under development. These potentials are combinable with the earlier tungsten-carbon-hydrogen ones. The potentials were used to explain the chemical sputtering of beryllium due to deuterium plasma exposure. During experiments, a large fraction of the sputtered beryllium atoms were observed to be released as BeD molecules, and the simulations identified the swift chemical sputtering mechanism, previously not believed to be important in metals, as the underlying mechanism. Radiation damage in the reactor structural materials vanadium, iron and iron chromium, as well as in the wall material tungsten and the mixed alloy tungsten carbide, was also studied in this thesis. Interatomic potentials for vanadium, tungsten and iron were modified to be better suited for simulating collision cascades that are formed during particle irradiation, and the potential features affecting the resulting primary damage were identified. Including the often neglected electronic effects in the simulations was also shown to have an impact on the damage. With proper tuning of the electronphonon interaction strength, experimentally measured quantities related to ion-beam mixing in iron could be reproduced. The damage in tungsten carbide alloys showed elemental asymmetry, as the major part of the damage consisted of carbon defects. On the other hand, modelling the damage

Fissile fuel assembly for a sub-moderated nuclear reactor

International Nuclear Information System (INIS)

Millot, J.P.; Dejeux, Pol.; Alibran, Patrice.

1983-01-01

Each of the core assemblies is composed of a prismatic case made of a neutron absorbing material, inside which very long rods containing the fissile material are arranged parallel to the height of the case and according to a regular network in the straight sections of the case. At least one piece in a fertile material exposed to the neutrons emitted by the fissile material of the assembly is arranged on each one of the side faces of the case. The invention applies in particular to sub-moderated reactors, cooled and moderated by pressurized water [fr

Structure optimization of CFB reactor for moderate temperature FGD

Energy Technology Data Exchange (ETDEWEB)

Li, Yuan; Zhang, Jie; Zheng, Kai; You, Changfu [Tsinghua Univ., Beijing (China). Dept. of Thermal Engineering; Ministry of Education, Beijing (China). Key Lab. for Thermal Science and Power Engineering

2013-07-01

The gas velocity distribution, sorbent particle concentration distribution and particle residence time in circulating fluidized bed (CFB) reactors for moderate temperature flue gas desulfurization (FGD) have significant influence on the desulfurization efficiency and the sorbent calcium conversion ratio for sulfur reaction. Experimental and numerical methods were used to investigate the influence of the key reactor structures, including the reactor outlet structure, internal structure, feed port and circulating port, on the gas velocity distribution, sorbent particle concentration distribution and particle residence time. Experimental results showed that the desulfurization efficiency increased 5-10% when the internal structure was added in the CFB reactor. Numerical analysis results showed that the particle residence time of the feed particles with the average diameter of 89 and 9 {mu}m increased 40% and 17% respectively, and the particle residence time of the circulating particles with the average diameter of 116 {mu}m increased 28% after reactor structure optimization. The particle concentration distribution also improved significantly, which was good for improving the contact efficiency between the sorbent particles and SO{sub 2}. In addition, the optimization guidelines were proposed to further increase the desulfurization efficiency and the sorbent calcium conversion ratio.

Nitrogen Removal by Anammox Biofilm Column Reactor at Moderately Low Temperature

Directory of Open Access Journals (Sweden)

Tuty Emilia Agustina

2017-10-01

Full Text Available The anaerobic ammonium oxidation (anammox as a new biological approach for nitrogen removal has been considered to be more cost-effective compared with the combination of nitrification and denitrification process. However, the anammox bioreactors are mostly explored at high temperature (>300C in which temperature controlling system is fully required. This research was intended to develop and to apply anammox process for high nitrogen concentration removal at ambient temperature used for treating wastewater in tropical countries. An up-flow biofilm column reactor, which the upper part constructed with a porous polyester non-woven fabric material as a carrier to attach the anammox bacteria was operated without heating system. A maximum nitrogen removal rate (NRR of 1.05 kg-N m3 d-1 was reached in the operation days of 178 with a Total Nitrogen (TN removal efficiency of 74%. This showed the biofilm column anammox reactor was successfully applied to moderate high nitrogen removal from synthetic wastewater at moderately low temperature. Keywords: Anammox, biofilm column reactor, ambient temperature, nitrogen removal

Method of collecting helium cover gas for heavy water moderated reactor

International Nuclear Information System (INIS)

Miyamoto, Keiji; Ueda, Hiroshi.

1981-01-01

Purpose: To reduce the systematic facility cost in a heavy water moderated reactor by contriving the simplification of a helium cover gas collecting intake system. Method: A detachable low pressure metal tank and a neoprene balloon are prepared for a vacuum pump in a permanent vacuum drying facility. When all of the helium cover gas is collected from a heavy water moderated reactor, a large capacity of neoprene balloon capable of temporarily storing it under low pressure is connected to the exhaust of the vacuum pump. On the other hand, while the reactor is operating, a suitable amount of the low pressure tank or neoprene balloon is connected to the exhaust side of the pump, thereby regulating the pressure of the helium cover gas. When refeeding the cover gas, the balloon, with a large capacity for collecting and storing the cover gas is connected to the intake side of the pump. Thus, the pressure regulation, collection of all of the cover gas and refeeding of the cover gas can be conducted without using a high discharge pump and high pressure tank. (Kamimura, M.)

Core design study on reduced-moderation water reactors

International Nuclear Information System (INIS)

Hiroshi, Akie; Yoshihiro, Nakano; Toshihisa, Shirakawa; Tsutomu, Okubo; Takamichi, Iwamura

2002-01-01

The conceptual core design study of reduced-moderation water reactors (RMWRs) with tight-pitched MOX-fuelled lattice has been carried out at JAERI. Several different RMWR core concepts based on both BWR and PWR have been proposed. All the core concepts meet with the aim to achieve both a conversion ratio of 1.0 or larger and negative void reactivity coefficient. As one of these RMWR concepts, the ABWR compatible core is also proposed. Although the conversion ratio of this core is 1.0 and the void coefficient is negative, the discharge burn-up of the fuel was about 25 GWd/t. By adopting a triangular fuel pin lattice for the reduction of moderator volume fraction and modifying axial Pu enrichment distribution, it was aimed to extend the discharge burn-up of ABWR compatible type RMWR. By using a triangular fuel lattice of smaller moderator volume fraction, discharge burn-up of 40 GWd/t seems achievable, keeping the high conversion ratio and the negative void coefficient. (authors)

Offshoots from beryllium development programme

International Nuclear Information System (INIS)

Sharma, B.P.; Sinha, P.K.

1995-01-01

The paper briefly presents extraction and processing of beryllium metal as practiced in the beryllium facilities at Turbhe, New Bombay. These facilities have been set up to meet the indigenous requirements of the metal in space and nuclear science programmes. As offshoot of this beryllium development programme has been the development of a number of pyro and powder metallurgical equipment. Indigenous development of these pieces of equipment has been a professionally rewarding experience. Efforts are now on to promote these equipment for industrial use. (author). 6 refs., 6 figs., 2 tabs

Tritium retention in S-65 beryllium after 100 eV plasma exposure

Energy Technology Data Exchange (ETDEWEB)

Causey, R.A. [Sandia National Labs., Livermore, CA (United States); Longhurst, G.R. [Idaho National Engineering Laboratories, Idaho Falls, 83415 (United States); Harbin, W. [Los Alamos National Laboratories, Los Alamos, NM 87545 (United States)

1997-02-01

The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 x 10{sup 17} (D+T)/cm{sup 2} s up to about 3 x 10{sup 18} (D+T)/cm{sup 2} s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H{sub 2} gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 x 10{sup 17} (D+T)/cm{sup 2} at 373 K to a low of 1 x 10{sup 16} (D+T)/cm{sup 2} at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C=0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters. (orig.).

Tritium retention in S-65 beryllium after 100 eV plasma exposure

Science.gov (United States)

Causey, Rion A.; Longhurst, Glen R.; Harbin, Wally

1997-02-01

The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 × 10 17 ( D + T)/ cm2s up to about 3 × 10 18 ( D + T)/ cm2s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H 2 gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 × 10 17 ( D + T)/ cm2 at 373 K to a low of 1 × 10 16 ( D + T)/ cm2 at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C = 0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters.

Effect of grain size on the hardness and reactivity of plasma-sintered beryllium

International Nuclear Information System (INIS)

Kim, Jae-Hwan; Nakamichi, Masaru

2014-01-01

Beryllium and its intermetallic compounds have attracted great attention as promising neutron multipliers in fusion reactors. In this study, mechanical and chemical properties of fabricated plasma-sintered beryllium (PS-Be) with different grain-sizes are investigated. Density and hardness analysis results of the fabricated PS-Be samples infer that a smaller grain size in the sintered Be indicates higher porosity and hardness. Sintered Be with a large grain size exhibits better resistance toward oxidation at 1273 K in dry air and at 1073 K in Ar/1% H 2 O, since oxidation at the grain boundaries of the determines the rate. In contrast, at 1273 K in Ar/1% H 2 O, a catastrophic oxidation is indicated by the increase of weight of the samples and the generation of H 2 from the bulk Be

Fluorimetric method for determination of Beryllium

International Nuclear Information System (INIS)

Sparacino, N.; Sabbioneda, S.

1996-10-01

The old fluorimetric method for the determination of Beryllium, based essentially on the fluorescence of the Beryllium-Morine complex in a strongly alkaline solution, is still competitive and stands the comparison with more modern methods or at least three reasons: in the presence of solid or gaseous samples (powders), the times necessary to finalize an analytic determination are comparable since the stage of the process which lasts the longest is the mineralization of the solid particles containing Beryllium, the cost of a good fluorimeter is by far Inferior to the cost, e. g., of an Emission Spectrophotometer provided with ICP torch and magnets for exploiting the Zeeman effect and of an Atomic absorption Spectrophotometer provided with Graphite furnace; it is possible to determine, fluorimetrically, rather small Beryllium levels (about 30 ng of Beryllium/sample), this potentiality is more than sufficient to guarantee the respect of all the work safety and hygiene rules now in force. The study which is the subject of this publication is designed to the analysis procedure which allows one to reach good results in the determination of Beryllium, chiefly through the control and measurement of the interference effect due to the presence of some metals which might accompany the environmental samples of workshops and laboratories where Beryllium is handled, either at the pure state or in its alloys. The results obtained satisfactorily point out the merits and limits of this analytic procedure

Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

International Nuclear Information System (INIS)

Hirose, T.; Ogiwara, H.; Enoeda, M.; Akiba, M.

2007-01-01

Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 μm and 10 μm were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

Energy Technology Data Exchange (ETDEWEB)

Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Ogiwara, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Enoeda, M. [Naka Fusion Research Establishment, J.A.E.R.I., Japan Atomic Energy Research Institute, Naka-gun, Ibaraki-ken (Japan); Akiba, M. [Naka Fusion Institute, Japan Atomic Energy Agency, Naka, Ibaraki (Japan)

2007-07-01

Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 {mu}m and 10 {mu}m were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

An optimization study of peak thermal neutron flux in moderators of advanced repetitive pulse reactors

International Nuclear Information System (INIS)

Asaoka, Takumi; Watanabe, N.

1976-01-01

In achieving a high peak thermal neutron flux in hydrogenous moderators installed in repetitive pulse reactors, the core-moderator arrangement can play as much an important role as the moderator design itself. However, the effect of the former has not been adequately emphasized to date, while a rather extensive study has been made on the latter. The present study concerns with a core-moderator system parameter optimization for a repetitive accelerator pulsed fast reactor. The results have shown that small differences in the arrangement resulting from the optimizations of various parameters are significant and the effects can be summed up to give an increase in the peak thermal flux by a factor of about two. (auth.)

A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

International Nuclear Information System (INIS)

Gesh, Christopher J.

2004-01-01

The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

Preparation of beryllium hydride

International Nuclear Information System (INIS)

Bergeron, C.R.; Baker, R.W.

1975-01-01

Beryllium hydride of high bulk density, suitable for use as a component of high-energy fuels, is prepared by the pyrolysis, in solution in an inert solvent, of a ditertiary-alkyl beryllium. An agitator introduces mechanical energy into the reaction system, during the pyrolysis, at the rate of 0.002 to 0.30 horsepower per gallon of reaction mixture. (U.S.)

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

High-strength beryllium block

International Nuclear Information System (INIS)

Pinto, N.P.; Keith, G.H.

1977-01-01

Beryllium billets hot isopressed using fine powder of high purity have exceptionally attractive properties; average tensile ultimate, 0.2% offset yield strength and elongation are 590 MPa, 430 MPa and 4.0% respectively. Properties are attributed to the fine grain size (about 4.0 μm average diameter) and the relatively low levels of BeO present as fine, well-dispersed particles. Dynamic properties, e.g., fracture toughness, are similar to those of standard grade, high-purity beryllium. The modulus of beryllium is retained to very high stress levels, and the microyield stress or precision elastic limit is higher than for other grades, including instrument grades. Limited data for billets made from normal-purity fine powders show similar room temperature properties. (author)

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

Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

International Nuclear Information System (INIS)

Muenzer, W.A.

1978-01-01

A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

Exposure and genetics increase risk of beryllium sensitisation and chronic beryllium disease in the nuclear weapons industry.

Science.gov (United States)

Van Dyke, Michael V; Martyny, John W; Mroz, Margaret M; Silveira, Lori J; Strand, Matt; Cragle, Donna L; Tankersley, William G; Wells, Susan M; Newman, Lee S; Maier, Lisa A

2011-11-01

Beryllium sensitisation (BeS) and chronic beryllium disease (CBD) are caused by exposure to beryllium with susceptibility affected by at least one well-studied genetic host factor, a glutamic acid residue at position 69 (E69) of the HLA-DPβ chain (DPβE69). However, the nature of the relationship between exposure and carriage of the DPβE69 genotype has not been well studied. The goal of this study was to determine the relationship between DPβE69 and exposure in BeS and CBD. Current and former workers (n=181) from a US nuclear weapons production facility, the Y-12 National Security Complex (Oak Ridge, Tennessee, USA), were enrolled in a case-control study including 35 individuals with BeS and 19 with CBD. HLA-DPB1 genotypes were determined by PCR-SSP. Beryllium exposures were assessed through worker interviews and industrial hygiene assessment of work tasks. After removing the confounding effect of potential beryllium exposure at another facility, multivariate models showed a sixfold (OR 6.06, 95% CI 1.96 to 18.7) increased odds for BeS and CBD combined among DPβE69 carriers and a fourfold (OR 3.98, 95% CI 1.43 to 11.0) increased odds for those exposed over an assigned lifetime-weighted average exposure of 0.1 μg/m(3). Those with both risk factors had higher increased odds (OR 24.1, 95% CI 4.77 to 122). DPβE69 carriage and high exposure to beryllium appear to contribute individually to the development of BeS and CBD. Among workers at a beryllium-using facility, the magnitude of risk associated with either elevated beryllium exposure or carriage of DPβE69 alone appears to be similar.

Effect of transient heating loads on beryllium

International Nuclear Information System (INIS)

Kupriyanov, Igor B.; Porezanov, Nicolay P.; Nikolaev, Georgyi N.; Kurbatova, Liudmila A.; Podkovyrov, Vyacheslav L.; Muzichenko, Anatoliy D.; Zhitlukhin, Anatoliy M.; Khimchenko, Leonid N.; Gervash, Alexander A.

2014-01-01

Highlights: • We study the effect of transient plasma loads on beryllium erosion and surface microstructure. • Beryllium targets were irradiated by plasma streams with energy of 0.5–1 MJ/m 2 at ∼250 °C. • Under plasma loads 0.5–1 MJ/m 2 cracking of beryllium surface is rather slight. • Under 0.5 MJ/m 2 the mass loss of Be is no more than 0.2 g/m 2 shot and decreasing with shots number. • Under 1 MJ/m 2 maximum mass loss of beryllium was 3.7 g/m 2 shot and decreasing with shots number. - Abstract: Beryllium will be used as a plasma facing material for ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of ITER first wall. The results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility are presented. The Be/CuCrZr mock-ups were exposed to upto 100 shots by deuterium plasma streams with pulse duration of 0.5 ms at ∼250 °C and average heat loads of 0.5 and 1 MJ/m 2 . Experiments were performed at 250 °C. The evolution of surface microstructure and cracks morphology as well as beryllium mass loss are investigated under erosion process

The Cryogenic Properties of Several Aluminum-Beryllium Alloys and a Beryllium Oxide Material

Science.gov (United States)

Gamwell, Wayne R.; McGill, Preston B.

2003-01-01

Performance related mechanical properties for two aluminum-beryllium (Al-Be) alloys and one beryllium-oxide (BeO) material were developed at cryogenic temperatures. Basic mechanical properties (Le., ultimate tensile strength, yield strength, percent elongation, and elastic modulus were obtained for the aluminum-beryllium alloy, AlBeMetl62 at cryogenic [-195.5"C (-320 F) and -252.8"C (-423"F)I temperatures. Basic mechanical properties for the Be0 material were obtained at cyrogenic [- 252.8"C (-423"F)] temperatures. Fracture properties were obtained for the investment cast alloy Beralcast 363 at cryogenic [-252.8"C (-423"F)] temperatures. The AlBeMetl62 material was extruded, the Be0 material was hot isostatic pressing (HIP) consolidated, and the Beralcast 363 material was investment cast.

Development of a standard for calculation and measurement of the moderator temperature coefficient of reactivity in water-moderated power reactors

International Nuclear Information System (INIS)

Mosteller, R.D.; Hall, R.A.; Lancaster, D.B.; Young, E.H.; Gavin, P.H.; Robertson, S.T.

1998-01-01

The contents of ANS 19.11, the standard for ''Calculation and Measurement of the Moderator Temperature Coefficient of Reactivity in Water-Moderated Power Reactors,'' are described. The standard addresses the calculation of the moderator temperature coefficient (MTC) both at standby conditions and at power. In addition, it describes several methods for the measurement of the at-power MTC and assesses their relative advantages and disadvantages. Finally, it specifies a minimum set of documentation requirements for compliance with the standard

Fault detection of sensors in nuclear reactors using self-organizing maps

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Paulo Roberto; Tiago, Graziela Marchi [Instituto Federal de Educacao, Ciencia e Tecnologia de Sao Paulo (IFSP), Sao Paulo, SP (Brazil); Bueno, Elaine Inacio [Instituto Federal de Educacao, Ciencia e Tecnologia de Sao Paulo (IFSP), Guarulhos, SP (Brazil); Pereira, Iraci Martinez, E-mail: martinez@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

In this work a Fault Detection System was developed based on the self-organizing maps methodology. This method was applied to the IEA-R1 research reactor at IPEN using a database generated by a theoretical model of the reactor. The IEA-R1 research reactor is a pool type reactor of 5 MW, cooled and moderated by light water, and uses graphite and beryllium as reflector. The theoretical model was developed using the Matlab Guide toolbox. The equations are based in the IEA-R1 mass and energy inventory balance and physical as well as operational aspects are taken into consideration. In order to test the model ability for fault detection, faults were artificially produced. As the value of the maximum calibration error for special thermocouples is +- 0.5 deg C, it had been inserted faults in the sensor signals with the purpose to produce the database considered in this work. The results show a high percentage of correct classification, encouraging the use of the technique for this type of industrial application. (author)

Fault detection of sensors in nuclear reactors using self-organizing maps

International Nuclear Information System (INIS)

Barbosa, Paulo Roberto; Tiago, Graziela Marchi; Bueno, Elaine Inacio; Pereira, Iraci Martinez

2011-01-01

In this work a Fault Detection System was developed based on the self-organizing maps methodology. This method was applied to the IEA-R1 research reactor at IPEN using a database generated by a theoretical model of the reactor. The IEA-R1 research reactor is a pool type reactor of 5 MW, cooled and moderated by light water, and uses graphite and beryllium as reflector. The theoretical model was developed using the Matlab Guide toolbox. The equations are based in the IEA-R1 mass and energy inventory balance and physical as well as operational aspects are taken into consideration. In order to test the model ability for fault detection, faults were artificially produced. As the value of the maximum calibration error for special thermocouples is +- 0.5 deg C, it had been inserted faults in the sensor signals with the purpose to produce the database considered in this work. The results show a high percentage of correct classification, encouraging the use of the technique for this type of industrial application. (author)

Beryllium poisonings; Les intoxications par le beryllium

Energy Technology Data Exchange (ETDEWEB)

Alibert, S.

1959-03-15

This note reports a bibliographical study of beryllium toxicity. Thus, this bibliographical review addresses and outlines aspects and issues like aetiology, cases of acute poisoning (cutaneous manifestations, pulmonary manifestations), chronic poisoning (cutaneous, pulmonary and bone manifestations), excretion and localisation, and prognosis.

Thermal fatigue behavior of US and Russian grades of beryllium

International Nuclear Information System (INIS)

Watson, R.D.; Youchison, D.L.; Dombrowski, D.E.; Guiniatouline, R.N.; Kupriynov, I.B.

1996-01-01

A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium which is proposed as plasma facing armor for fusion reactor first wall, limiter, and divertor components. The 30 KW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature ''spike'' of 750 degrees C for each pass of the beam. Large thermal stress in excess of the yield strength are generated due to very high spot heat flux, 250 MW/m 2 . Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S-65H, S-200F, S-300F-H, Sr-200, I-400, extruded high purity. HIP'd sperical powder, porous beryllium (94% and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe 12 . Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation, and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT, and TShG-56. Rolled sheet Be(SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis

The impact of transient thermal loads on beryllium as plasma facing material

Energy Technology Data Exchange (ETDEWEB)

Spilker, Benjamin Christof

2017-01-24

The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO{sub 2} free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m{sup -2} range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby

The impact of transient thermal loads on beryllium as plasma facing material

International Nuclear Information System (INIS)

Spilker, Benjamin Christof

2017-01-01

The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO_2 free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m"-"2 range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby, the

The immunotoxicity of beryllium

International Nuclear Information System (INIS)

Reeves, A.L.

1983-01-01

In the disease berylliosis, granulomatous hypersensitivity is the specific immune response to tissue contact with a poorly soluble particle of beryllium compound, mediated through the accumulation and proliferation of reticuloendothelial cells. A review is given of the work accomplished since the 1950's and particularly since the 1970's to elucidate the nature and consequences of this response to beryllium and its compounds. (U.K.)

Use of a Paraffin Based Grout to Stabilize Buried Beryllium and Other Wastes

International Nuclear Information System (INIS)

Gretchen Matthern; Duane Hanson; Neal Yancey; Darrell Knudson

2005-01-01

The long term durability of WAXFIXi, a paraffin based grout, was evaluated for in situ grouting of activated beryllium wastes in the Subsurface Disposal Area (SDA), a radioactive landfill at the Radioactive Waste Management Complex, part of the Idaho National Laboratory (INL). The evaluation considered radiological and biological mechanisms that could degrade the grout using data from an extensive literature search and previous tests of in situ grouting at the INL. Conservative radioactive doses for WAXFIX were calculated from the ''hottest'' (i.e., highest-activity) Advanced Test Reactor beryllium block in the SDA.. These results indicate that WAXFIX would not experience extensive radiation damage for many hundreds of years. Calculation of radiation induced hydrogen generation in WAXFIX indicated that grout physical performance should not be reduced beyond the effects of radiation dose on the molecular structure. Degradation of a paraffin-based grout by microorganisms in the SDA is possible and perhaps likely, but the rate of degradation will be at a slower rate than found in the literature reviewed. The calculations showed the outer 0.46 m (18 in.) layer of each monolith, which represents the minimum expected distance to the beryllium block, was calculated to require 1,000 to 3,600 years to be consumed. The existing data and estimations of biodegradation and radiolysis rates for WAXFIX/paraffin do not indicate any immediate problems with the use of WAXFIX for grouting beryllium or other wastes in the SDA

Investigation of beryllium/steam interaction

Energy Technology Data Exchange (ETDEWEB)

Chekhonadskikh, A.M.; Vurim, A.D.; Vasilyev, Yu.S.; Pivovarov, O.S. [Inst. of Atomic Energy National Nuclear Center of the Republic of Kazakstan Semipalatinsk (Kazakhstan); Shestakov, V.P.; Tazhibayeva, I.L.

1998-01-01

In this report program on investigations of beryllium emissivity and transient processes on overheated beryllium surface attacked by water steam to be carried out in IAE NNC RK within Task S81 TT 2096-07-16 FR. The experimental facility design is elaborated in this Report. (author)

Assessment of LANL beryllium waste management documentation

International Nuclear Information System (INIS)

Danna, J.G.; Jennrich, E.A.; Lund, D.M.; Davis, K.D.; Hoevemeyer, S.S.

1991-04-01

The objective of this report is to determine present status of the preparation and implementation of the various high priority documents required to properly manage the beryllium waste generated at the Laboratory. The documents being assessed are: Waste Acceptance Criteria, Waste Characterization Plan, Waste Certification Plan, Waste Acceptance Procedures, Waste Characterization Procedures, Waste Certification Procedures, Waste Training Procedures and Waste Recordkeeping Procedures. Beryllium is regulated (as a dust) under 40 CFR 261.33 as ''Discarded commercial chemical products, off specification species, container residues and spill residues thereof.'' Beryllium is also identified in the 3rd thirds ruling of June 1, 1990 as being restricted from land disposal (as a dust). The beryllium waste generated at the Laboratory is handled separately because beryllium has been identified as a highly toxic carcinogenic material

Chronic Beryllium Disease Prevention Program Report

Energy Technology Data Exchange (ETDEWEB)

Lee, S

2012-03-29

This document describes how Lawrence Livermore National Laboratory (LLNL) meets the requirements and management practices of federal regulation 10 CFR 850, 'Chronic Beryllium Disease Prevention Program (CBDPP).' This revision of the LLNL CBDPP incorporates clarification and editorial changes based on lessons learned from employee discussions, observations and reviews of Department of Energy (DOE) Complex and commercial industry beryllium (Be) safety programs. The information is used to strengthen beryllium safety practices at LLNL, particularly in the areas of: (1) Management of small parts and components; and (2) Communication of program status to employees. Future changes to LLNL beryllium activities and on-going operating experience will be incorporated into the program as described in Section S, 'Performance Feedback.'

Status of research and development on reduced-moderation water reactors

International Nuclear Information System (INIS)

Iwamura, Takamichi

2002-01-01

To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

Status of research and development on reduced-moderation water reactors

Energy Technology Data Exchange (ETDEWEB)

Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

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

Beryllium concentration in pharyngeal tonsils in children

Directory of Open Access Journals (Sweden)

Ewa Nogaj

2014-06-01

Full Text Available Power plant dust is believed to be the main source of the increased presence of the element beryllium in the environment which has been detected in the atmospheric air, surface waters, groundwater, soil, food, and cigarette smoke. In humans, beryllium absorption occurs mainly via the respiratory system. The pharyngeal tonsils are located on the roof of the nasopharynx and are in direct contact with dust particles in inhaled air. As a result, the concentration levels of beryllium in the pharyngeal tonsils are likely to be a good indicator of concentration levels in the air. The presented study had two primary aims: to investigate the beryllium concentration in pharyngeal tonsils in children living in southern Poland, and the appropriate reference range for this element in children’s pharyngeal tonsils. Pharyngeal tonsils were extracted from a total of 379 children (age 2–17 years, mean 6.2 ± 2.7 years living in southern Poland. Tonsil samples were mineralized in a closed cycle in a pressure mineralizer PDS 6, using 65% spectrally pure nitric acid. Beryllium concentration was determined using the ICP-AES method with a Perkin Elmer Optima 5300DVTM. The software Statistica v. 9 was used for the statistical analysis. It was found that girls had a significantly greater beryllium concentration in their pharyngeal tonsils than boys. Beryllium concentration varies greatly, mostly according to the place of residence. Based on the study results, the reference value for beryllium in pharyngeal tonsils of children is recommended to be determined at 0.02–0.04 µg/g.

New audio applications of beryllium metal

International Nuclear Information System (INIS)

Sato, M.

1977-01-01

The major applications of beryllium metal in the field of audio appliances are for the vibrating cones for the two types of speakers 'TWITTER' for high range sound and 'SQUAWKER' for mid range sound, and also for beryllium cantilever tube assembled in stereo cartridge. These new applications are based on the characteristic property of beryllium having high ratio of modulus of elasticity to specific gravity. The production of these audio parts is described, and the audio response is shown. (author)

Design and development of rolled joint for moderator sparger channel of an Indian Pressurised Heavy Water Reactor

International Nuclear Information System (INIS)

Joemon, V.; Sinha, R.K.

1993-01-01

Indian Pressurised Heavy Water Reactors are natural uranium fuelled heavy water moderated and cooled reactors. As per the conventional scheme, the moderator enters through one or more inlet nozzles penetrating the calandria shell and flows out through outlet nozzles. Baffles are fixed at the inlet nozzles for proper distribution of moderator in the calandria and to avoid the impact of the jet on the neighbouring calandria tubes. An alternate scheme for moderator inlet has been conceived and engineered in which three lower peripheral lattice locations of the reactor are converted into moderator inlets. This is achieved by moderator sparger channels each containing a 5 m long perforated zircaloy-2 sparger tube rolled to the calandria tube sheets and extended by stainless steel tubular components (inserts) at both ends of a sparger channel. Moderator enters the sparger channel at both ends and flows into the calandria. In the absence of standard codes for design of rolled joints, it was requires to develop these joints based on trials followed by various tests. this paper discusses the details of the rolled joint developed for this purpose, the details of the trials with test results and optimization of rolling parameters for these joints

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

Material and geometry options and performance characteristics for a test reactor

International Nuclear Information System (INIS)

Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

1993-01-01

For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

Energy Technology Data Exchange (ETDEWEB)

Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

2013-05-15

To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

Study on thermal neutron spectra in reactor moderators by time-of-flight method

International Nuclear Information System (INIS)

Akino, Fujiyoshi

1982-12-01

Prediction of thermal neutron spectra in a reactor core plays very important role in the neutronic design of the reactor for obtaining the accurate thermal group constants. It is well known that the neutron scattering properties of the moderator materials markedly influence the thermal neutron spectra. Therefore, 0 0 angular dependent thermal neutron spectra were measured by the time-of-flight method in the following moderator bulks 1) Graphite bulk poisoned with boron at the temperatures from 20 to 800 0 C, 2) Light water bulk poisoned with Cadmium and/or Indium, 3) Light water-natural uranium heterogeneous bulk. The measured results were compared with calculation utilizing Young-Koppel and Haywood scattering model for graphite and light water respectively. On the other hand, a variety of 20% enriched uranium loaded and graphite moderated cores consisting of the different lattice cell in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments related to Very High Temperature Reactor (VHTR). The experimental data were for the critical masses in 235 U, reactivity worths of experimental burnable poison rods, thorium rods, natural-uranium rods and experimental control rods and kinetic parameters. It is made clear from comparison between measurement and calculation that the accurate thermal group constants can be obtained by use of the Young-Koppel and Haywood neutron scattering models if heterogeneity of reactor core lattices is taken into account precisely. (author)

Beryllium and graphite performance in ITER during a disruption

International Nuclear Information System (INIS)

Hassanein, A.; Ehst, D.A.; Gahl, J.

1994-01-01

Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed. ((orig.))

Beryllium and graphite performance in ITER during a disruption

International Nuclear Information System (INIS)

Hassanein, A.; Ehst, D.A.; Gahl, J.

1993-09-01

Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed

BERYLLIUM MEASUREMENT IN COMMERCIALLY AVAILABLE WET WIPES

Energy Technology Data Exchange (ETDEWEB)

Youmans-Mcdonald, L.

2011-02-18

Analysis for beryllium by fluorescence is now an established method which is used in many government-run laboratories and commercial facilities. This study investigates the use of this technique using commercially available wet wipes. The fluorescence method is widely documented and has been approved as a standard test method by ASTM International and the National Institute for Occupational Safety and Health (NIOSH). The procedure involves dissolution of samples in aqueous ammonium bifluoride solution and then adding a small aliquot to a basic hydroxybenzoquinoline sulfonate fluorescent dye (Berylliant{trademark} Inc. Detection Solution Part No. CH-2) , and measuring the fluorescence. This method is specific to beryllium. This work explores the use of three different commercial wipes spiked with beryllium, as beryllium acetate or as beryllium oxide and subsequent analysis by optical fluorescence. The effect of possible interfering metals such as Fe, Ti and Pu in the wipe medium is also examined.

Beryllium Measurement In Commercially Available Wet Wipes

International Nuclear Information System (INIS)

Youmans-Mcdonald, L.

2011-01-01

Analysis for beryllium by fluorescence is now an established method which is used in many government-run laboratories and commercial facilities. This study investigates the use of this technique using commercially available wet wipes. The fluorescence method is widely documented and has been approved as a standard test method by ASTM International and the National Institute for Occupational Safety and Health (NIOSH). The procedure involves dissolution of samples in aqueous ammonium bifluoride solution and then adding a small aliquot to a basic hydroxybenzoquinoline sulfonate fluorescent dye (Berylliant(trademark) Inc. Detection Solution Part No. CH-2) , and measuring the fluorescence. This method is specific to beryllium. This work explores the use of three different commercial wipes spiked with beryllium, as beryllium acetate or as beryllium oxide and subsequent analysis by optical fluorescence. The effect of possible interfering metals such as Fe, Ti and Pu in the wipe medium is also examined.

Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

Energy Technology Data Exchange (ETDEWEB)

Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

2014-08-15

IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

Beryllium dust generation resulting from plasma bombardment

International Nuclear Information System (INIS)

Doerner, R.; Mays, C.

1997-01-01

The beryllium dust resulting from erosion of beryllium samples subjected to plasma bombardment has been measured in PISCES-B. Loose surface dust was found to be uniformly distributed throughout the device and accounts for 3% of the eroded material. A size distribution measurement of the loose surface dust shows an increasing number of particles with decreasing diameter. Beryllium coatings on surfaces with a line of sight view of the target interaction region account for an additional 33% of the eroded beryllium material. Flaking of these surface layers is observed and is thought to play a significant role in dust generation inside the vacuum vessel. (orig.)

Technical issues for beryllium use in fusion blanket applications

International Nuclear Information System (INIS)

McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

1985-01-01

Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented

Production of beryllium oxide of nuclear purity from beryl

Energy Technology Data Exchange (ETDEWEB)

Copat, A; Sood, S P

1984-01-01

Production of beryllium oxide from beryl by the fluoride process was optimized in this study. Optimum results were obtained using a mixture of sodium hexafluorsilicate and sodium hexafluorferrate as flux and calcinating at 740/sup 0/C for 2 hours. The beryllium concentrate produced was further purified by crystallization as beryllium sulfate to obtain nuclear grade beryllium oxide

Production of beryllium oxide of nuclear purity from beryl

International Nuclear Information System (INIS)

Copat, A.; Sood, S.P.

1983-01-01

Production of beryllium oxide from beryl by the fluoride process was optimized in this study. Optimum results were obtained using a mixture of sodium hexafluorsilicate and sodium hexafluorferrate as flux and calcinating at 740 0 C for 2 hours. The beryllium concentrate produced was further purified by crystallization as beryllium sulfate to obtain nuclear grade beryllium oxide (Author) [pt

Hydrogen isotopes transport parameters in fusion reactor materials

International Nuclear Information System (INIS)

Serra, E.; Ogorodnikova, O.V.

1998-01-01

This work presents a review of hydrogen isotopes-materials interactions in various materials of interest for fusion reactors. The relevant parameters cover mainly diffusivity, solubility, trap concentration and energy difference between trap and solution sites. The list of materials includes the martensitic steels (MANET, Batman and F82H-mod.), beryllium, aluminium, beryllium oxide, aluminium oxide, copper, tungsten and molybdenum. Some experimental work on the parameters that describe the surface effects is also mentioned. (orig.)

Preliminary results for explosion bonding of beryllium to copper

International Nuclear Information System (INIS)

Butler, D.J.; Dombrowski, D.E.

1995-01-01

This program was undertaken to determine if explosive bonding is a viable technique for joining beryllium to copper substrates. The effort was a cursory attempt at trying to solve some of the problems associated with explosive bonding beryllium and should not be considered a comprehensive research effort. There are two issues that this program addressed. Can beryllium be explosive bonded to copper substrates and can the bonding take place without shattering the beryllium? Thirteen different explosive bonding iterations were completed using various thicknesses of beryllium that were manufactured with three different techniques

Preliminary proposal for a beryllium technology program for fusion applications

International Nuclear Information System (INIS)

1985-02-01

The program was designed to provide the answers to the critical issues of beryllium technology needed in fusion blanket designs. The four tasks are as follows: (1) Beryllium property measurements needed for fusion data base. (2) Beryllium stress relaxation and creep measurements for lifetime modelling calculations. (3) Simplified recycle technique development for irradiated beryllium. (4) Beryllium neutron multiplier measurements using manganese bath absolute calibration techniques

Ion beam assisted deposition of metal-coatings on beryllium

International Nuclear Information System (INIS)

Tashlykov, I.S.; Tul'ev, V.V.

2015-01-01

Thin films were applied on beryllium substrates on the basis of metals (Cr, Ti, Cu and W) with method of the ion-assisted deposition in vacuum. Me/Be structures were prepared using 20 kV ions irradiation during deposition on beryllium neutral fraction generated from vacuum arc plasma. Rutherford back scattering and computer simulation RUMP code were applied to investigate the composition of the modified beryllium surface. Researches showed that the superficial structure is formed on beryllium by thickness ~ 50-60 nm. The covering composition includes atoms of the deposited metal (0.5-3.3 at. %), atoms of technological impurity carbon (0.8-1.8 at. %) and oxygen (6.3-9.9 at. %), atoms of beryllium from the substrate. Ion assisted deposition of metals on beryllium substrate is accompanied by radiation enhanced diffusion of metals, oxygen atoms in the substrate, out diffusion of beryllium, carbon atoms in the deposited coating and sputtering film-forming ions assists. (authors)

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.

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

1994-01-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

The moderator's moderator

International Nuclear Information System (INIS)

Williamson, G.K.

1990-01-01

A brief account is given of the development of graphite moderators for Magnox and advanced gas cooled reactors. The accident at Windscale in 1957 brought to worldwide attention the importance of irradiation damage in graphite and the consequent storage of Wigner energy. In spite of the Windscale setback, preparations for the civil programme of Magnox reactors went ahead apace. Some of the background to the disastrous Dungeness B tender is presented. In spite of all the difficulties and uncertainties, the graphite in UK reactors has performed well. In all cases, as far as the author is aware, the behaviour of the graphite moderators will not prevent design life being achieved. (author)

CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

Energy Technology Data Exchange (ETDEWEB)

Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

2015-10-15

Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

Beryllium-aluminum alloys for investment castings

International Nuclear Information System (INIS)

Nachtrab, W.T.; Levoy, N.

1997-01-01

Beryllium-aluminum alloys containing greater than 60 wt % beryllium are very favorable materials for applications requiring light weight and high stiffness. However, when produced by traditional powder metallurgical methods, these alloys are expensive and have limited applications. To reduce the cost of making beryllium-aluminum components, Nuclear Metals Inc. (NMI) and Lockheed Martin Electronics and Missiles have recently developed a family of patented beryllium-aluminum alloys that can be investment cast. Designated Beralcast, the alloys can achieve substantial weight savings because of their high specific strength and stiffness. In some cases, weight has been reduced by up to 50% over aluminum investment casting. Beralcast is now being used to make thin wall precision investment castings for several advanced aerospace applications, such as the RAH-66 Comanche helicopter and F-22 jet fighter. This article discusses alloy compositions, properties, casting method, and the effects of cobalt additions on strength

Characterization of plasma sprayed beryllium ITER first wall mockups

Energy Technology Data Exchange (ETDEWEB)

Castro, R.G.; Vaidya, R.U.; Hollis, K.J. [Los Alamos National Lab., NM (United States). Material Science and Technology Div.

1998-01-01

ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/m{sup 2} without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface. (author)

Characterization of Plasma Sprayed Beryllium ITER First Wall Mockups

International Nuclear Information System (INIS)

Castro, Richard G.; Vaidya, Rajendra U.; Hollis, Kendall J.

1997-10-01

ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/sq m without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface

Reference Monte Carlo calculations of Maria reactor core

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

2002-01-01

The reference Monte Carlo calculations of MARIA reactor core have been carried to evaluate accuracy of the calculations at each stage of its neutron-physics analysis using deterministic codes. The elementary cell has been calculated with two main goals; evaluation of effects of simplifications introduced in deterministic lattice spectrum calculations by the WIMS code and evaluation of library data in recently developed WIMS libraries. In particular the beryllium data of those libraries needed evaluation. The whole core calculations mainly the first MARIA critical experiment and the first critical core after the 8-year break in operation. Both cores contained only fresh fuel elements but only in the first critical core the beryllium blocks were not poisoned by Li-6 and He-3. Thus the MCNP k-eff results could be compared with the experiment. The MCNP calculations for the cores with beryllium poisoned suffered the deficiency of uncertainty in the poison concentration, but a comparison of power distribution shows that realistic poison levels have been carried out for the operating reactor MARIA configurations. (author)

Modernization and Refurbishment of the RECH-1 Reactor

Energy Technology Data Exchange (ETDEWEB)

Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

2014-08-15

The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

Implanted Deuterium Retention and Release in Carbon-Coated Beryllium

Science.gov (United States)

Anderl, R. A.; Longhurst, G. R.; Pawelko, R. J.; Oates, M. A.

1997-06-01

Deuterium implantation experiments have been conducted on samples of clean and carbon-coated beryllium. These studies entailed preparation and characterization of beryllium samples coated with carbon thicknesses of 100, 500, and 1000 Å. Heat treatment of a beryllium sample coated with carbon to a thickness of approximately 100 Å revealed that exposure to a temperature of 400°C under high vacuum conditions was sufficient to cause substantial diffusion of beryllium through the carbon layer, resulting in more beryllium than carbon at the surface. Comparable concentrations of carbon and beryllium were observed in the bulk of the coating layer. Higher than expected oxygen levels were observed throughout the coating layer as well. Samples were exposed to deuterium implantation followed by thermal desorption without exposure to air. Differences were observed in deuterium retention and postimplantation release behavior in the carbon-coated samples as compared with bare samples. For comparable implantation conditions (sample temperature of 400°C and an incident deuterium flux of approximately 6 × 1019 D/m2-s), the quantity of deuterium retained in the bare sample was less than that retained in the carbon-coated samples. Further, the release of the deuterium took place at lower temperatures for the bare beryllium surfaces than for carbon-coated beryllium samples.

MEASUREMENTS OF THE PROPERTIES OF BERYLLIUM FOIL

International Nuclear Information System (INIS)

ZHAO, Y.; WANG, H.

2000-01-01

The electrical conductivity of beryllium at radio frequency (800 MHz) and liquid nitrogen temperature were investigated and measured. This summary addresses a collection of beryllium properties in the literature, an analysis of the anomalous skin effect, the test model, the experimental setup and improvements, MAFIA simulations, the measurement results and data analyses. The final results show that the conductivity of beryllium is not as good as indicated by the handbook, yet very close to copper at liquid nitrogen temperature

Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

Energy Technology Data Exchange (ETDEWEB)

Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

2002-12-01

The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)

Measurement of cold neutron spectra at a model of cryogenic moderator of the IBR-2M reactor

International Nuclear Information System (INIS)

Kulikov, S.A.; Chernikov, A.N.; Shabalin, E.P.; Kalinin, I.V.; Morozov, V.M.; Novikov, A.G.; Puchkov, A.V.

2010-01-01

The article is dedicated to methods and results of experimental determination of cold neutron spectra from solid mesitylene at neutron moderator temperatures 10-50 K. Experiments were fulfilled at the DIN-2PI spectrometer of the IBR-2 reactor. The main goals of this work were to examine a system of constants for Monte Carlo calculation of cryogenic moderators of the IBR-2M reactor and to determine the temperature dependence of cold neutron intensity from the moderator. A reasonable agreement of experimental and calculation results for mesitylene at 20 K has been obtained. The cold neutron intensity at temperature of moderator 10 K is about 1.8 times higher than at T=50 K

Heavy water moderated reactors advances and challenges

International Nuclear Information System (INIS)

Meneley, D.A.; Olmstead, R.A.; Yu, A.M.; Dastur, A.R.; Yu, S.K.W.

1994-01-01

Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

Preparation of a sinterable beryllium oxide through decomposition of beryllium hydroxide (1963)

International Nuclear Information System (INIS)

Bernier, M.

1963-01-01

In the course of the present study, we have attempted to precise the factors which among the ones effective in the course of the preparation of the beryllium hydroxide and oxide and during the sintering have an influence on the final result: the density and homogeneity of the sintered body. Of the several varieties of hydroxides precipitated from a sulfate solution the β-hydroxide only is always contaminated with beryllium sulfate and cannot be purified even by thorough washing. We noticed that those varieties of the hydroxide (gel, α, β) have different decomposition rates; this behaviour is used to identify and even to dose the different species in (α, β) mixtures. The various hydroxides transmit to the resulting oxides the shape they had when precipitated. Accordingly the history of the oxide is revealed by its behaviour during its fabrication and sintering. By comparing the results of the sintering operation with the various measurements performed on the oxide powders we are led to the conclusion that an oxide obtained from beryllium hydroxide is sinterable under vacuum if the following conditions are fulfilled: the particle size must lie between 0.1 and 0.2 μ and the BeSO 4 content of the powder must be less than 0.25 per cent wt (expressed as SO 3 /BeO). The best fitting is obtained with the oxide issued from an α-hydroxide precipitated as very small aggregates and with a low sulfur-content. We have observed that this is also the case for the oxide obtained by direct calcination of beryllium sulfate. (author) [fr

Reference moderator calculated performance for the LANSCE upgrade project

International Nuclear Information System (INIS)

Ferguson, P.D.; Russell, G.J.; Pitcher, E.J.

1995-01-01

The authors have calculated the performance of five moderators of interest to the LANSCE upgrade project. Coupled and decoupled light water and liquid hydrogen moderators in flux-trap geometry surrounded by a neutronically infinite heavy-water cooled beryllium reflector have been studied. Time and energy spectra, as well as semi-empirical fits to the data, are presented. The data has been made available to aid the instrument design and moderator selection process

Beryllium poisonings

International Nuclear Information System (INIS)

Alibert, S.

1959-03-01

This note reports a bibliographical study of beryllium toxicity. Thus, this bibliographical review addresses and outlines aspects and issues like aetiology, cases of acute poisoning (cutaneous manifestations, pulmonary manifestations), chronic poisoning (cutaneous, pulmonary and bone manifestations), excretion and localisation, and prognosis

Effect of scaling on the thermal hydraulics of the moderator of a CANDU reactor

International Nuclear Information System (INIS)

Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

2011-01-01

Three dimensional numerical simulations are conducted on the CANDU Moderator Test Facility (MTF) and the actual size CANDU reactor. Moderator test facility is ¼ scale of the actual reactor. The heat input and other operating conditions are scaled down from the real reactor to the MTF using constant Archimedes number (as considered in MTF experiments performed by Atomic Energy of Canada Ltd.). The heat generations inside both tanks are applied through volumetric heating. In this method, heat is added to the fluid throughout the volume as it occurs in real reactor through fission heat generation and gamma rays from radioactive materials. The temperatures in actual reactor simulation are about 10 deg C greater than in MTF simulations. The separation between high and low temperature zones are more visible in real reactor simulation comparing to MTF simulation. The result indicates that the MTF has better mixing and weaker buoyancy forces comparing to real reactor. The velocity distribution in both cases seems similar with impingement point for inlet jets in both cases at the right hand side of the tank. Although the velocities are considerably higher (about 40%) in the case of real reactor, but as we go toward inner core of the tanks, the velocities are similar and very low. Several points inside the tank are monitored for their temperature and velocity with time. The results for these points show fluctuations in both temperature and velocity inside the tank. The fluctuations frequency seems higher in the case of real reactor while the amplitude of fluctuations is smaller in real reactor in most of the points. Here, in this research we have shown that Archimedes number alone cannot be a good scaling parameter (as used in MTF experiments) and it should be used along with Rayleigh number for scaling purposes. (author)

Effect of machining damage on tensile properties of beryllium

International Nuclear Information System (INIS)

Hanafee, J.E.

1976-01-01

It is well established that damage introduced at the surface of beryllium during machining operations can lower its mechanical properties. Tensile tests were conducted to illustrate this on beryllium presently being used for parts in the W79 program and similar to the new powder-processed beryllium specified for production (tentative specification MEL 76-001319). The objective of this study is to quantitatively illuminate the importance of controlling machining damage in this particular grade of powder-processed beryllium

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Dien, Nguyen Nhi; Lam, Pham Van; Phuong, Huynh Dong; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10{sup 5}/10{sup 8} n/cm{sup 2}/sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to ({gamma},n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is {beta}{sup B}e{sub eff}=0.49%{beta}{sub eff} for a beryllium weight relative to U{sup 235} fuel of m{sub B}e/m{sub U} = 8.5. This result is acceptable in comparison to those obtained for other Be-U{sup 235} media. (author). 5 refs., 2 figs., 4 tabs.

Kinetics of Pressurized Water Reactors with Hot or Cold Moderators

Energy Technology Data Exchange (ETDEWEB)

Norinder, O

1960-11-15

The set of neutron kinetic equations developed in this report permits the use of long integration steps during stepwise integration. Thermal relations which describe the transfer of heat from fuel to coolant are derived. The influence upon the kinetic behavior of the reactor of a number of parameters is studied. A comparison of the kinetic properties of the hot and cold moderators is given.

Sintering of beryllium oxide

International Nuclear Information System (INIS)

Caillat, R.; Pointud, R.

1955-01-01

This study had for origin to find a process permitting to manufacture bricks of beryllium oxide of pure nuclear grade, with a density as elevated as possible and with standardized shape. The sintering under load was the technique kept for the manufacture of the bricks. Because of the important toxicity of the beryllium oxide, the general features for the preliminary study of the sintering, have been determined while using alumina. The obtained results will be able to act as general indication for ulterior studies with sintering under load. (M.B.) [fr

Pelletized cold moderator of the IBR-2 reactor: current status and future development

International Nuclear Information System (INIS)

Ananiev, V; Beliakov, A; Bulavin, M; Verkhogliadov, A; Kulagin, E; Kulikov, S; Mukhin, K; Shabalin, E; Loktaev, K

2016-01-01

Current status and future development of the pelletized cold moderator of the IBR-2 reactor in Neutron Physics Laboratory of JINR are represented. Nowadays cold moderator works for physical experiments and allows conducting experiments in the region of wavelengths more than 4 Å up to 10-13 times faster in comparison with the warm water moderator. Future development of the pelletized cold moderator is aimed at increasing the time of its operation for experiments and is based on three components: creation of a system of continuous charging and discharging of beads, supplementation of various additives, and use of new materials, such as triphenylmethane. (paper)

Feasibility study of a fission supressed blanket for a tandem-mirror hybrid reactor

International Nuclear Information System (INIS)

Moir, R.W.; Lee, J.D.; Barr, W.L.

1981-01-01

A study of fission suppressed blankets for the tandem mirror not only showed such blankets to be feasible but also to be safer than fissioning blankets. Such hybrids could produce enough fissile material to support up to 17 light water reactors of the same nuclear power rating. Beryllium was compared to 7 Li for neutron multiplication; both were considered feasible but the blanket with Li produced 20% less fissile fuel per unit of nuclear power in the reactor. The beryllium resource, while possibly being too small for extensive pure fusion application, would be adequate (with carefully planned industrial expansion) for the hybrid because of the large support ratio, and hence few hybrids required. Radiation damage and coatings for beryllium remain issues to be resolved by further study and experimentation. Molten salt reprocessing was compared to aqueous solution reprocessing

Managing health effects of beryllium exposure

National Research Council Canada - National Science Library

Committee on Beryllium Alloy Exposures; Committee on Toxicology; National Research Council; Division on Earth and Life Studies; National Research Council

2008-01-01

... to its occurrence in exposed people. Despite reduced workplace exposure, chronic beryllium disease continues to occur. In addition, beryllium has been classified as a likely human carcinogen by several agencies, such as the International Agency for Research on Cancer, the National Toxicology Program, and the U.S. Environmental Protection Agency. Thos...

Influence of neutron irradiation on the tritium retention in beryllium

Energy Technology Data Exchange (ETDEWEB)

Rolli, R.; Ruebel, S.; Werle, H. [Forschungszentrum Karlsruhe, Inst. fuer Neutronenphysik und Reaktortechnik, Karlsruhe (Germany); Wu, C.H.

1998-01-01

Carbon-based materials and beryllium are the candidates for protective layers on the components of fusion reactors facing plasma. In contact with D-T plasma, these materials absorb tritium, and it is anticipated that tritium retention increases with the neutron damage due to neutron-induced traps. Because of the poor data base for beryllium, the work was concentrated on it. Tritium was loaded into the samples from stagnant T{sub 2}/H{sub 2} atmosphere, and afterwards, the quantity of the loaded tritium was determined by purged thermal annealing. The specification of the samples is shown. The samples were analyzed by SEM before and after irradiation. The loading and the annealing equipments are contained in two different glove boxes with N{sub 2} inert atmosphere. The methods of loading and annealing are explained. The separation of neutron-produced and loaded tritium and the determination of loaded tritium in irradiated samples are reported. Also the determination of loaded tritium in unirradiated samples is reported. It is evident that irradiated samples contained much more loaded tritium than unirradiated samples. The main results of this investigation are summarized in the table. (K.I.)

Low enriched uranium UAl{sub X}-Al targets for the production of Molybdenum-99 in the IEA-R1 and RMB reactors

Energy Technology Data Exchange (ETDEWEB)

Domingos, Douglas B.; Silva, Antonio T. e; Joao, Thiago G.; Silva, Jose Eduardo R. da, E-mail: teixeira@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Nishiyama, Pedro J.B. de O., E-mail: pedro.julio@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

2011-07-01

The IEA-R1 reactor of IPEN/CNEN-SP in Brazil is a pool type research reactor cooled and moderated by demineralized water and having Beryllium and Graphite as reflectors. In 1997 the reactor received the operating licensing for 5 MW. A new research reactor is being planned in Brazil to replace the IEA-R1 reactor. This new reactor, the Brazilian Multipurpose Reactor (RMB), planned for 30 MW, is now in the conception design phase. Low enriched uranium (LEU) (<20% {sup 235}U) UAl{sub x} dispersed in Al targets are being considered for production of Molybdenum-99 ({sup 99}Mo) by fission. Neutronic and thermal-hydraulics calculations were performed, respectively, to compare the production of {sup 99}Mo for these targets in IEA-R1 reactor and RMB and to determine the temperatures achieved in the UAl{sub x}-Al targets during irradiation. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION, CITATION and SCALE and for the thermal-hydraulics calculations was utilized the computer code MTRCR-IEAR1. (author)

Implanted deuterium retention and release in carbon-coated beryllium

International Nuclear Information System (INIS)

Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Oates, M.A.

1997-01-01

Deuterium implantation experiments have been conducted on samples of clean and carbon-coated beryllium. These studies entailed preparation and characterization of beryllium samples coated with carbon thicknesses of 100, 500, and 1000 angstrom. Heat treatment of a beryllium sample coated with carbon to a thickness of approximately 100 angstrom revealed that exposure to a temperature of 400 degrees C under high vacuum conditions was sufficient to cause substantial diffusion of beryllium through the carbon layer, resulting in more beryllium than carbon at the surface. Comparable concentrations of carbon and beryllium were observed in the bulk of the coating layer. Higher than expected oxygen levels were observed throughout the coating layer as well. Samples were exposed to deuterium implantation followed by thermal desorption without exposure to air. Differences were observed in deuterium retention and postimplantation release behavior in the carbon-coated samples as compared with bare samples. For comparable implantation conditions (sample temperature of 400 degrees C and an incident deuterium flux of approximately 6 X 10 19 D/m 2 sec), the quantity of deuterium retained in the bare sample was less than that retained in the carbon-coated samples. Further, the release of the deuterium took place at lower temperatures for the bare beryllium surfaces than for carbon-coated beryllium samples. 4 refs., 8 figs., 1 tab

Beryllium in aircraft brakes - a summary

International Nuclear Information System (INIS)

Zenczak, S.

1977-01-01

Beryllium has been in use in aircraft brakes for ten years. During the original design phases of the several aircraft programs using beryllium a number of problems requiring solution confronted the designers. In actual service the solution to these problems performed much better than had been anticipated. A summary is presented. (author)

Galvanic corrosion of beryllium welds

International Nuclear Information System (INIS)

Hill, M.A.; Butt, D.P.; Lillard, R.S.

1997-01-01

Beryllium is difficult to weld because it is highly susceptible to cracking. The most commonly used filler metal in beryllium welds is Al-12 wt.% Si. Beryllium has been successfully welded using Al-Si filler metal with more than 30 wt.% Al. This filler creates an aluminum-rich fusion zone with a low melting point that tends to backfill cracks. Drawbacks to adding a filler metal include a reduction in service temperature, a lowering of the tensile strength of the weld, and the possibility for galvanic corrosion to occur at the weld. To evaluate the degree of interaction between Be and Al-Si in an actual weld, sections from a mock beryllium weldment were exposed to 0.1 M Cl - solution. Results indicate that the galvanic couple between Be and the Al-Si weld material results in the cathodic protection of the weld and of the anodic dissolution of the bulk Be material. While the cathodic protection of Al is generally inefficient, the high anodic dissolution rate of the bulk Be during pitting corrosion combined with the insulating properties of the Be oxide afford some protection of the Al-Si weld material. Although dissolution of the Be precipitate in the weld material does occur, no corrosion of the Al-Si matrix was observed

The liquid hydrogen moderator at the NIST research reactor

International Nuclear Information System (INIS)

Williams, Robert E.; Rowe, J. Michael; Kopetka, Paul

1997-09-01

In 1995, the NIST research reactor was shut down for a number of modifications, including the replacement of the D 2 O cold neutron source with a liquid hydrogen moderator. When the liquid hydrogen source began operating, the flux of cold neutrons increased by a factor of six over the D 2 O source. The design and operation of the hydrogen source are described, and measurements of its performance are compared with the Monte Carlo simulations used in the design. (auth)

Research on Reduced-Moderation Water Reactor (RMWR)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro

1999-11-01

The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from 238 U to 239 Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

Research on Reduced-Moderation Water Reactor (RMWR)

Energy Technology Data Exchange (ETDEWEB)

Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

1999-11-01

The Reduced-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor which aims at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. These characteristics can be achieved by the high conversion ratio from {sup 238}U to {sup 239}Pu resulted from the higher neutron energy spectrum in comparison to conventional light water reactors. Considering the extension of LWR utilization, Japan Atomic Energy Research Institute (JAERI) started the research on it in 1997 and then started a collaboration in the conceptual design study with the Japan Atomic Power Company (JAPCO) in 1998. In the core design study of the RMWR, negative void reactivity coefficient is required from a viewpoint of safety as well as establishing hard neutron spectrum. In order to achieve the above trade-off characteristics simultaneously, several basic core design ideas should be combined, such as a tight lattice fuel assembly, a flat core, a blanket effect, a streaming effect and so on. Up to now, five core concepts have been created for the RMWR as follows: a high conversion BWR with high void fraction and super-flat core, a long operation cycle BWR using void channels, a high conversion BWR without blankets, a high conversion PWR using heavy water as a coolant, and a PWR for plutonium multi-recycle using seed-blanket type fuel assemblies. The present report summarizes the objectives, domestic and international trends, principles and characteristics, core conceptual designs and future R and D plans of the RMWR. (J.P.N.)

The 10 MW multipurpose TRIGA reactor at Ongkharak Nuclear Research Center, Thailand

International Nuclear Information System (INIS)

Thurgood, B.E.; Razvi, J.; Whittemore, J.L.; Bhadrakom, K.

1997-01-01

General Atomics (GA), has been selected to lead a team of firms from the United States, Japan, Australia and Thailand to design, build and commission the Ongkharak Nuclear Research Center near Bangkok, Thailand, for the Office of Atomic Energy for Peace. The facilities to be provided comprise of: A Reactor Island, consisting of a 10 MW TRIGA reactor that takes full advantage of the inherent safety characteristics of uranium-zirconium hydride (UZrH) fuel; An Isotope Production Facility for the production of radioisotopes and radiopharmaceuticals using the TRIGA reactor; A Waste Processing and Storage Facility for the processing and storage of radioactive waste from the facility as well as other locations in Thailand. The centerpiece of the Center will be the TRIGA reactor, fueled with low-enriched UZrH fuel, cooled and moderated by light water, and reflected by beryllium and heavy water. The UZrH fueled reactor will have a rated steady state thermal power output of 10 MW, and will be capable of performing the following: Radioisotope production for medical, industrial and agricultural uses; Neutron transmutation doping of silicon; Beam experiments such as Neutron Scattering, Neutron Radiography (NR), and Prompt Gamma Neutron Activation Analysis (PGNAA); Medical therapy of patients using Boron Neutron Capture Therapy (BNCT); Applied research and technology development in the nuclear field; Training in principles of reactor operation, reactor physics, reactor experiments, etc. (author)

MD simulation: determination of the physical properties and surface vaporization analysis of beryllium armours

International Nuclear Information System (INIS)

Prinzio, M. Di; Aquaro, D.

2006-01-01

The erosion of the divertor and of the first wall determined on the base of the anticipated operating conditions, is a critical issue that could affect the performance and the operating schedule of the nuclear fusion reactor ITER. This paper deals with the analysis of beryllium thermal properties by means of MD simulations, in order to better predict thermal behaviour of beryllium armoured PFCs in fusion devices. The importance of this analysis is clearly connected to thermal response evaluation of PFCs to high heat flux exposure, during off-normal events and Edge Localized Modes. The ensuing strong over-heating, in fact, produces material ablation through vaporization of surface material layers and possible loss of melting material. The overall PFCs erosion has bearings on plasma contamination, due to eroded material transport, and components lifetime, due to armour thickness reduction. An important feature of beryllium is its high vapour pressure. During thermal transients the strong vaporization keeps surface temperature relatively low but eroded thickness results high as well. Small changes in beryllium vapour pressure produce not negligible differences in thermal analyses results. On the basis of available force fields, classical Molecular Dynamics simulations have been carried out in order to better understand surface vaporization in tokamak conditions and to evaluate the effect of beryllium oxides formation. This effect has been successfully modelled by MD simulation, carried out with Moldy code. Morse stretching and bending potential for Be-O bond simulation have been used, and partial charges method, accounting for molecular polarity, has been employed. Since during short thermal transients, such as ELMs, only a few microns of Be armour will be overheated and reach melting threshold, the effective thermal conductivity is very important in determining the temperature evolution of surface layers and the ensuing erosion. Thermal conductivity can be evaluated

Structure investigations of some beryllium materials

Energy Technology Data Exchange (ETDEWEB)

Faeldt, I; Lagerberg, G

1960-05-15

Metallographic structure, microhardness and texture have been studied on various types of beryllium metal including hot pressed powder, a rolled strip and an extruded tube It was found that beryllium exhibits its highest hardness in directions perpendicular to the basal plane. Good ideas of the prevailing textures were obtained with an ordinary X-ray diffractometer.

Structure investigations of some beryllium materials

International Nuclear Information System (INIS)

Faeldt, I.; Lagerberg, G.

1960-05-01

Metallographic structure, microhardness and texture have been studied on various types of beryllium metal including hot pressed powder, a rolled strip and an extruded tube It was found that beryllium exhibits its highest hardness in directions perpendicular to the basal plane. Good ideas of the prevailing textures were obtained with an ordinary X-ray diffractometer

The adhesive bonding of beryllium structural components

International Nuclear Information System (INIS)

Fullerton-Batten, R.C.

1977-01-01

Where service conditions permit, adhesive bonding is a highly recommendable, reliable means of joining beryllium structural parts. Several important programs have successfully used adhesive bonding for joining structural and non-structural beryllium components. Adhesive bonding minimizes stress concentrations associated with other joining techniques and considerably improves fatigue resistance. In addition, no degradation of base metal properties occur. In many instances, structural joints can be fabricated more cheaply by adhesive bonding or in combination with adhesive bonding than by any other method used alone. An evaluation program on structural adhesive bonding of beryllium sheet components is described. A suitable surface pretreatment for beryllium adherends prior to bonding is given. Tensile shear strength and fatigue properties of FM 1000 and FM 123-5 adhesive bonded joints are reviewed and compared with data obtained from riveted joints of similar geometry. (author)

Beryllium armour produced by evaporation-condensation technique

International Nuclear Information System (INIS)

Anisimov, A.; Frolov, V.; Moszherin, S.; Pepekin, G.; Pirogov, A.; Komarov, V.; Mazul, I.

1997-01-01

Beryllium, as armour material for ITER plasma facing components, has a limited erosion lifetime. In order to repair the surface of eroded tiles in-situ, Be-deposition technologies are under consideration. One of them uses the physical vapour deposition of beryllium on copper or beryllium substrate produced by a hot Be-target placed in the vicinity of this substrate. Three different options for using this technology for ITER Be-armour application are considered. The first option is the repair in-situ of eroded Be-tiles. The second option suggests the use of this technology to provide the joining of Be to Cu-substrate. The third option assumes the use of evaporated-condensed beryllium as a bulk tile material bonded to copper substrate by conventional joining (Brazing et al.) techniques. The first results and prospects of these approaches are presented below. (orig.)

Use of a Paraffin Based Grout to Stabilize Buried Beryllium and Other Wastes

Energy Technology Data Exchange (ETDEWEB)

Gretchen Matthern; Duane Hanson; Neal Yancey; Darrell Knudson

2005-12-01

The long term durability of WAXFIXi, a paraffin based grout, was evaluated for in situ grouting of activated beryllium wastes in the Subsurface Disposal Area (SDA), a radioactive landfill at the Radioactive Waste Management Complex, part of the Idaho National Laboratory (INL). The evaluation considered radiological and biological mechanisms that could degrade the grout using data from an extensive literature search and previous tests of in situ grouting at the INL. Conservative radioactive doses for WAXFIX were calculated from the "hottest" (i.e., highest-activity) Advanced Test Reactor beryllium block in the SDA.. These results indicate that WAXFIX would not experience extensive radiation damage for many hundreds of years. Calculation of radiation induced hydrogen generation in WAXFIX indicated that grout physical performance should not be reduced beyond the effects of radiation dose on the molecular structure. Degradation of a paraffin-based grout by microorganisms in the SDA is possible and perhaps likely, but the rate of degradation will be at a slower rate than found in the literature reviewed. The calculations showed the outer 0.46 m (18 in.) layer of each monolith, which represents the minimum expected distance to the beryllium block, was calculated to require 1,000 to 3,600 years to be consumed. The existing data and estimations of biodegradation and radiolysis rates

Advanced concept of reduced-moderation water reactor (RMWR) for plutonium multiple recycling

International Nuclear Information System (INIS)

Okubo, T.; Iwamura, T.; Takeda, R.; Yamamoto, K.; Okada, H.

2001-01-01

An advanced water-cooled reactor concept named the Reduced-Moderation Water Reactor (RMWR) has been proposed to attain a high conversion ratio more than 1.0 and to achieve the negative void reactivity coefficient. At present, several types of design concepts satisfying both the design targets have been proposed based on the evaluation for the fuel without fission products and minor actinides. In this paper, the feasibility of the RMWR core is investigated for the plutonium multiple recycling under advanced reprocessing schemes with low decontamination factors as proposed for the FBR fuel cycle. (author)

Metallurgical viewpoints on the brittleness of beryllium

Energy Technology Data Exchange (ETDEWEB)

Lagerberg, G

1960-02-15

At present the development and use of beryllium metal for structural applications is severely hampered by its brittleness. Reasons for this lack of ductility are reviewed in discussing the deformation behaviour of beryllium in relation to other hexagonal metals. The ease of fracturing in beryllium is assumed to be a consequence of a limited number of deformation modes in combination with high deformation resistance. Models for the nucleation of fracture are suggested. The relation of ductility to elastic constants as well as to grain size, texture and alloying additions is discussed.

Metallurgical viewpoints on the brittleness of beryllium

International Nuclear Information System (INIS)

Lagerberg, G.

1960-02-01

At present the development and use of beryllium metal for structural applications is severely hampered by its brittleness. Reasons for this lack of ductility are reviewed in discussing the deformation behaviour of beryllium in relation to other hexagonal metals. The ease of fracturing in beryllium is assumed to be a consequence of a limited number of deformation modes in combination with high deformation resistance. Models for the nucleation of fracture are suggested. The relation of ductility to elastic constants as well as to grain size, texture and alloying additions is discussed

Laser welding of a beryllium/tantalum collimator

International Nuclear Information System (INIS)

Lingenfelter, A.C.; Anglin, C.D.

1985-01-01

This report describes the methods utilized in the fabrication of a collimator from 0.001 inch thick beryllium and tantalum foil. The laser welding process proved to be an acceptable method for joining the beryllium in a standing edge joint configuration

Beryllium brazing considerations in CANDU fuel bundle manufacture

International Nuclear Information System (INIS)

Harmsen, J.; Pant, A.; Lewis, B.J.; Thompson, W.T.

2010-01-01

'Full text:' Appendages of CANDU fuel bundle elements are currently joined to zircaloy sheaths by vacuum beryllium brazing. Ongoing environmental and workplace concerns about beryllium combined with the continuous efforts by Cameco Fuel Manufacturing in its improvement process, initiated this study to find a substitute for pure beryllium. The presentation will review the necessary functionality of brazing alloy components and short list a series of alloys with the potential to duplicate the performance of pure beryllium. Modifications to current manufacturing processes based on in-plant testing will be discussed in relation to the use of these alloys. The presentation will conclude with a summary of the progress to date and further testing expected to be necessary.

Deuterium permeation and diffusion in high-purity beryllium

International Nuclear Information System (INIS)

Abramov, E.; Riehm, M.P.; Thompson, D.A.; Smeltzer, W.W.

1990-01-01

The permeation rate of deuterium through high-purity beryllium membranes was measured using the gas-driven permeation technique. The time-dependent and the steady-state deuterium flux data were analyzed and the effective diffusivities of the samples were determined. Using multilayer permeation theory the effects of surface oxide were eliminated and the diffusion coefficients of the bulk beryllium determined. The diffusion parameters obtained for the extra-grade beryllium samples (99.8%) are D 0 =6.7x10 -9 m 2 /s and E D =28.4 kJ/mol. For the high-grade beryllium samples (99%) the parameters are D 0 =8.0x10 -9 m 2 /s and E D =35.1 kJ/mol. (orig.)

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

1992-01-01

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

An automatic regulating control system for a graphite moderated reactor using digital techniques

International Nuclear Information System (INIS)

Carvalho Goncalves Junior, J. de.

1989-01-01

The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

Advantages of liquid fluoride thorium reactor in comparison with light water reactor

Energy Technology Data Exchange (ETDEWEB)

Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M. [Nuclear Science Program, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia)

2015-04-29

Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

Quantitative method of determining beryllium or a compound thereof in a sample

Science.gov (United States)

McCleskey, T. Mark; Ehler, Deborah S.; John, Kevin D.; Burrell, Anthony K.; Collis, Gavin E.; Minogue, Edel M.; Warner, Benjamin P.

2010-08-24

A method of determining beryllium or a beryllium compound thereof in a sample, includes providing a sample suspected of comprising beryllium or a compound thereof, extracting beryllium or a compound thereof from the sample by dissolving in a solution, adding a fluorescent indicator to the solution to thereby bind any beryllium or a compound thereof to the fluorescent indicator, and determining the presence or amount of any beryllium or a compound thereof in the sample by measuring fluorescence.

Bioenvironmental Engineering Guide to Beryllium

Science.gov (United States)

2017-07-26

Dermal contact with beryllium can result in dermatitis resembling first- or second-degree burns and skin granulomas [7]. Beryllium dust, fume...minute short-term exposure limit (STEL) of 2.0 µg/m3 [§1910.1024(c)(2) & §1926.1124(c)(2)], and added provisions to prevent skin contact [§1910.1024(b...document you want more information, contact the Environmental, Safety, and Occupational Health (ESOH) Service Center at DSN 798-3764, 1-888-232-ESOH (3764

Preparation of a sinterable beryllium oxide through decomposition of beryllium hydroxide (1963); Preparation d'un oxyde de beryllium frittable par decomposition de l'hydiloxyde (1963)

Energy Technology Data Exchange (ETDEWEB)

Bernier, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

In the course of the present study, we have attempted to precise the factors which among the ones effective in the course of the preparation of the beryllium hydroxide and oxide and during the sintering have an influence on the final result: the density and homogeneity of the sintered body. Of the several varieties of hydroxides precipitated from a sulfate solution the {beta}-hydroxide only is always contaminated with beryllium sulfate and cannot be purified even by thorough washing. We noticed that those varieties of the hydroxide (gel, {alpha}, {beta}) have different decomposition rates; this behaviour is used to identify and even to dose the different species in ({alpha}, {beta}) mixtures. The various hydroxides transmit to the resulting oxides the shape they had when precipitated. Accordingly the history of the oxide is revealed by its behaviour during its fabrication and sintering. By comparing the results of the sintering operation with the various measurements performed on the oxide powders we are led to the conclusion that an oxide obtained from beryllium hydroxide is sinterable under vacuum if the following conditions are fulfilled: the particle size must lie between 0.1 and 0.2 {mu} and the BeSO{sub 4} content of the powder must be less than 0.25 per cent wt (expressed as SO{sub 3}/BeO). The best fitting is obtained with the oxide issued from an {alpha}-hydroxide precipitated as very small aggregates and with a low sulfur-content. We have observed that this is also the case for the oxide obtained by direct calcination of beryllium sulfate. (author) [French] Au cours de cette etude, nous avons cherche a preciser les facteurs qui, intervenant tout au long de la preparation de l'hydroxyde, puis de l'oxyde de beryllium et enfin du frittage, peuvent avoir une influence sur le resultat final: la densite et l'homogeneite du fritte. Parmi tous les hydroxydes precipites d'une solution de sulfate, seul l'hydroxyde {beta} est toujours fortement pollue par le sulfate

Impurities effect on the swelling of neutron irradiated beryllium

International Nuclear Information System (INIS)

Donne, M.D.; Scaffidi-Argentina, F.

1995-01-01

An important factor controlling the swelling behaviour of fast neutron irradiated beryllium is the impurity content which can strongly affect both the surface tension and the creep strength of this material. Being the volume swelling of the old beryllium (early sixties) systematically higher than that of the more modem one (end of the seventies), a sensitivity analysis with the aid of the computer code ANFIBE (ANalysis of Fusion Irradiated BEryllium) to investigate the effect of these material properties on the swelling behaviour of neutron irradiated beryllium has been performed. Two sets of experimental data have been selected: the first one named Western refers to quite recently produced Western beryllium, whilst the second one, named Russian refers to relatively old (early sixties) Russian beryllium containing a higher impurity rate than the Western one. The results obtained with the ANFIBE Code were assessed by comparison with experimental data and the used material properties were compared with the data available in the literature. Good agreement between calculated and measured values has been found

Gaseous-fuel nuclear reactor research for multimegawatt power in space

Science.gov (United States)

Thom, K.; Schneider, R. T.; Helmick, H. H.

1977-01-01

In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

Modeling of hydrogen interactions with beryllium

Energy Technology Data Exchange (ETDEWEB)

Longhurst, G.R. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

1998-01-01

In this paper, improved mathematical models are developed for hydrogen interactions with beryllium. This includes the saturation effect observed for high-flux implantation of ions from plasmas and retention of tritium produced from neutronic transmutations in beryllium. Use of the models developed is justified by showing how they can replicated experimental data using the TMAP4 tritium transport code. (author)

Assessment of irradiation effects on beryllium reflector and heavy water tank of JRR-3M

Energy Technology Data Exchange (ETDEWEB)

Murayama, Yoji; Kakehuda, Kazuhiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

The JRR-3M, a swimming pool type research reactor with beryllium and heavy water reflectors, has been operated since 1990. Since the beryllium reflectors are close to fuel and receive high fast neutron fluence in a relatively short time, they may be subject to change their dimensions by swelling due mostly to entrapped helium gaseous. This may bend the reflectors to the outside and narrow gaps between the reflectors and the fuel elements. The gaps have been measured with an ultrasonic thickness gage in an annual inspection. The results in 1996 show that the maximum of expansion in the diametral directions was 0.6 mm against 1.6 mm of a managed value for replacement of the reflector. A heavy water tank of the JRR-3M is made of aluminum alloy A5052. Surveillance tests of the alloy have been conducted to evaluate irradiation effects of the heavy water tank. Five sets of specimens of the alloy have been irradiated in the beryllium reflectors where fast neutron flux is higher than that in the heavy water tank. In 1994, one set of specimens had been unloaded and carried out the post-irradiation tests. The results show that the heavy water tank preserved satisfactory mechanical properties. (author)

Deuterium permeation and diffusion in high purity beryllium

International Nuclear Information System (INIS)

Abramov, E.

1990-05-01

The permeation rate of deuterium through high-purity beryllium membranes was measured using the gas-driven permeation technique. The time-dependent and the steady-state deuterium flux data were analyzed and the effective diffusivities of the samples were determined. A multilayer permeation theory was used in order to eliminate the surface oxide effects and the diffusion coefficients of the bulk beryllium were determined. The diffusion parameters obtained for the extra-grade beryllium samples (99.8%) are D 0 = 6.7 x 10 -9 [m 2 /s] and E D = 28.4 [KJ/mol]; and for the high-grade beryllium samples (99%) the parameters are D 0 = 8.0 x 10 -9 [m 2 /s] and E D = 35.1 [KJ/mol

Formation of cellular structure in beryllium at plastic working

International Nuclear Information System (INIS)

Papirov, I.I.; Nikolaenko, A.A.; Shokurov, V.S.; Pikalov, A.I.

2013-01-01

Conditions of cellular structure formation are investigated at various kinds of deformation and heat treatment of beryllium ingots. It is shown that the cellular structure plays the important role in formation of complex of physical mechanical properties of beryllium. Influence of impurity, various conditions of deformation (temperature, squeezing degree) and heat treatments on substructure, texture and mechanical properties of metal is investigated. Optimum conditions of rolling and heat treatments of beryllium are defined. The way of sign-variable cyclic deformation of beryllium ingots is offered for reception quasi-isotropic fine-grained metal. Physical-mechanical properties of ultra fine-grained metal are studied

Upgrading and modernization of the high flux reactor Petten

International Nuclear Information System (INIS)

Ahlf, J.

1992-01-01

The High Flux Reactor (HFR) at Petten, The Netherlands, owned by the European Communities and operated by the Netherlands Energy Research Foundation, is a water-cooled and moderated, multipurpose research reactor of the closed-tank in-pool type, operated at 45 MW. Performance upgrading comprised two power increases from 20 MW via 30 MW to 45 MW, providing more and higher rated irradiation positions in the tank. With the replacement of the original reactor vessel the experimental capabilities of the reactor were improved. Better pool side facilities and the introduction of a large cross-section, double, beam tube were implemented. Additional major installation upgrading activities consisted of the replacement of the primary and the pool heat exchangers, replacement of the beryllium reflector elements, extension of the overpower protection systems and upgrading of the nuclear instrumentation as well as the guaranteed power supply. Control room upgrading is in progress. A full new safety analysis, as well as the introduction of a comprehensive Quality Assurance system, are summarized under software upgrading. Continuous modernization and upgrading also takes place of equipment for fuel and structural materials irradiations for fission reactors and future fusion machines. In parallel, all supporting services, as well as the management structure for large irradiation programmes, have been developed. Presently the reactor is operating at about 275 full power days per year with an average utilization of the irradiation positions of 70 to 80%. (orig.)

Low cycle thermal fatigue testing of beryllium grades for ITER plasma facing components

International Nuclear Information System (INIS)

Watson, R.D.; Youchison, D.L.; Dombrowski, D.E.; Guiniatouline, R.N.; Kupriynov, I.B.

1996-01-01

A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium, which is proposed as plasma facing armor for fusion reactor first wall, limiter, and divertor components. The 30 kW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature ''spike'' of 750 degree C for each pass of the beam. Large thermal stresses in excess of the yield strength are generated due to very high spot heat flux, 250 MW/m 2 . Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S- 65H, S-200F, S-200F-H, SR-200, I-400, extruded high purity, HIP'd spherical powder, porous beryllium (94% and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe 12 . Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation, and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT, and TShG-56. Rolled sheet Be (SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis

Preparation of copper-beryllium alloys from Indian beryl

International Nuclear Information System (INIS)

Paul, C.M.; Sharma, B.P.; Subba Rao, K.S.; Rajadhyaksha, M.G.; Sundaram, C.V.

1975-01-01

The paper presents the results of laboratory-scale investigations on the preparation of copper-beryllium and aluminium beryllium master alloys starting from Indian beryl and adopting the fluoride process. The flowsheet involves: (1) conversion of the Be-values in beryl into water soluble sodium beryllium fluoride, (2) preparation of beryllium hydroxide by alkali treatment of aqueous Na 2 BeF 4 (3) conversion of Be(OH) 2 to (NH 4 ) 2 BeF 4 by treatment with NH 4 HF 2 (4) thermal decomposition of (NH 4 ) 2 BeF 4 to BeF 2 and (5) magnesium reduction of BeF 2 (without/with) the addition of copper/aluminium to obtain beryllium metal/alloys. The method has been successfully employed for the preparation of Cu-Be master alloys containing about 8% Be and free of Mg on a 200 gm scale. A1-80% Be master alloys have also been prepared by this method. Toxicity and health hazards associated with Be are discussed and the steps taken to ensure safe handling of Be are described. (author)

10 CFR 850.20 - Baseline beryllium inventory.

Science.gov (United States)

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Baseline beryllium inventory. 850.20 Section 850.20 Energy... Baseline beryllium inventory. (a) The responsible employer must develop a baseline inventory of the... inventory, the responsible employer must: (1) Review current and historical records; (2) Interview workers...

Characterization of shocked beryllium

Directory of Open Access Journals (Sweden)

Papin P.A.

2012-08-01

Full Text Available While numerous studies have investigated the low-strain-rate constitutive response of beryllium, the combined influence of high strain rate and temperature on the mechanical behavior and microstructure of beryllium has received limited attention over the last 40 years. In the current work, high strain rate tests were conducted using both explosive drive and a gas gun to accelerate the material. Prior studies have focused on tensile loading behavior, or limited conditions of dynamic strain rate and/or temperature. Two constitutive strength (plasticity models, the Preston-Tonks-Wallace (PTW and Mechanical Threshold Stress (MTS models, were calibrated using common quasi-static and Hopkinson bar data. However, simulations with the two models give noticeably different results when compared with the measured experimental wave profiles. The experimental results indicate that, even if fractured by the initial shock loading, the Be remains sufficiently intact to support a shear stress following partial release and subsequent shock re-loading. Additional “arrested” drive shots were designed and tested to minimize the reflected tensile pulse in the sample. These tests were done to both validate the model and to put large shock induced compressive loads into the beryllium sample.

Interaction of implanted deuterium and helium with beryllium: radiation enhanced oxidation

International Nuclear Information System (INIS)

Langley, R.A.

1979-01-01

The interaction of implanted deuterium and helium with beryllium is of significant interest in the application of first wall coatings and other components of fusion reactors. Electropolished polycrystalline beryllium was first implanted with an Xe backscatter marker at 1.98 MeV followed by either implantation with 5 keV diatomic deuterium or helium. A 2.0 MeV He beam was used to analyze for impurity buildup; namely oxygen. The oxide layer thickness was found to increase linearly with increasing implant fluence. A 2.5 MeV H + beam was used to depth profile the D and He by ion backscattering. In addition the retention of the implant was measured as a function of the implant fluence. The mean depth of the implant was found to agree with theoretical range calculations. Scanning electron microscopy was used to observe blister formation. No blisters were observed for implanted D but for implanted He blisters occurred at approx. 1.75 x 10 17 He cm -2 . The blister diameter increased with increasing implant fluence from about 0.8 μm at 10 18 He cm -2 to 5.5 μm at 3 x 10 18 He cm -2

Photoluminescence enhancement from GaN by beryllium doping

Science.gov (United States)

García-Gutiérrez, R.; Ramos-Carrazco, A.; Berman-Mendoza, D.; Hirata, G. A.; Contreras, O. E.; Barboza-Flores, M.

2016-10-01

High quality Be-doped (Be = 0.19 at.%) GaN powder has been grown by reacting high purity Ga diluted alloys (Be-Ga) with ultra high purity ammonia in a horizontal quartz tube reactor at 1200 °C. An initial low-temperature treatment to dissolve ammonia into the Ga melt produced GaN powders with 100% reaction efficiency. Doping was achieved by dissolving beryllium into the gallium metal. The powders synthesized by this method regularly consist of two particle size distributions: large hollow columns with lengths between 5 and 10 μm and small platelets in a range of diameters among 1 and 3 μm. The GaN:Be powders present a high quality polycrystalline profile with preferential growth on the [10 1 bar 1] plane, observed by means of X-ray diffraction. The three characteristics growth planes of the GaN crystalline phase were found by using high resolution TEM microscopy. The optical enhancing of the emission in the GaN powder is attributed to defects created with the beryllium doping. The room temperature photoluminescence emission spectra of GaN:Be powders, revealed the presence of beryllium on a shoulder peak at 3.39 eV and an unusual Y6 emission at 3.32eV related to surface donor-acceptor pairs. Also, a donor-acceptor-pair transition at 3.17 eV and a phonon replica transition at 3.1 eV were observed at low temperature (10 K). The well-known yellow luminescence band coming from defects was observed in both spectra at room and low temperature. Cathodoluminescence emission from GaN:Be powders presents two main peaks associated with an ultraviolet band emission and the yellow emission known from defects. To study the trapping levels related with the defects formed in the GaN:Be, thermoluminescence glow curves were obtained using UV and β radiation in the range of 50 and 150 °C.

Occupational and non-occupational allergic contact dermatitis from beryllium

Energy Technology Data Exchange (ETDEWEB)

Vilaplana, J; Romaguera, C; Grimalt, F [Allergy Department, Dermatological Service Hospital Clinico, Barcelona (Spain)

1992-01-01

There are various references to sensitization to beryllium in the literature. Since introducing a patch testing series for patients with suspected sensitization to metals, we have found 3 cases of sensitization to beryllium. Of these 3 cases, we regard the first 2 as having relevant sensitization. Beryllium chloride (1% pet.) was positive in 3 patients and negative in 150 controls. (au).

Occupational and non-occupational allergic contact dermatitis from beryllium

International Nuclear Information System (INIS)

Vilaplana, J.; Romaguera, C.; Grimalt, F.

1992-01-01

There are various references to sensitization to beryllium in the literature. Since introducing a patch testing series for patients with suspected sensitization to metals, we have found 3 cases of sensitization to beryllium. Of these 3 cases, we regard the first 2 as having relevant sensitization. Beryllium chloride (1% pet.) was positive in 3 patients and negative in 150 controls. (au)

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2000-01-01

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

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2000-07-01

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

Calculation of the Thermal State of the Graphite Moderator of the RBMK Reactor

Directory of Open Access Journals (Sweden)

Vorobiev Alexander V.

2017-01-01

Full Text Available This work is devoted to study the temperature field of the graphite stack of the RBMK reactor. In work was analyzed the influence of contact pressure between the components of the masonry on the temperature of the graphite moderator.

Moderator feedback effects in two-dimensional nodal methods for pressurized water reactor analysis

International Nuclear Information System (INIS)

Downar, T.J.

1987-01-01

A method was developed for incorporating moderator feedback effects in two-dimensional nodal codes used for pressurized water reactor (PWR) neutronic analysis. Equations for the assembly average quality and density are developed in terms of the assembly power calculated in two dimensions. The method is validated with a Westinghouse PWR using the Electric Power Research Institute code SIMULATE-E. Results show a several percent improvement is achieved in the two-dimensional power distribution prediction compared to methods without moderator feedback

Deuterium/hydrogen isotope exchange on beryllium and beryllium nitride; Deuterium/Wasserstoff-Isotopenaustausch an Beryllium und Berylliumnitrid

Energy Technology Data Exchange (ETDEWEB)

Dollase, Petra; Eichler, Michael; Koeppen, Martin; Dittmar, Timo; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik (Germany)

2016-07-01

In the fusion experiments JET and ITER, the first wall is made up of beryllium. The use of nitrogen is discussed for radiative cooling in the divertor. This can react with the surface of the first wall to form beryllium nitride (Be{sub 3}N{sub 2}). The hydrogen isotopes deuterium and tritium, which react in the fusion reaction to helium and a neutron, are used as fuel. Since the magnetic confinement of the plasma is not perfect, deuterium and tritium ions are also found on the beryllium wall and can accumulate there. This should be avoided due to the radioactivity of tritium. Therefore the isotope exchange with deuterium is investigated to regenerate the first wall. We investigate the isotopic exchange of deuterium and protium in order to have not to work with radioactive tritium. The ion bombardment is simulated with an ion source. With voltages up to a maximum of 5 kV, deuterium and protic hydrogen ions are implanted in polycrystalline Be and Be{sub 3}N{sub 2}. The samples are then analyzed in situ using X-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS). Subsequently, samples prepared under the same conditions are characterized ex-situ by means of nuclear reaction analysis (NRA). [German] In den Fusionsexperimenten JET und ITER besteht die erste Wand im Hauptraum aus Beryllium (Be). Zur Strahlungskuehlung im Divertor wird der Einsatz von Stickstoff diskutiert. Dieser kann mit der Oberflaeche der ersten Wand zu Berylliumnitrid (Be{sub 3}N{sub 2}) reagieren. Als Brennstoff werden die Wasserstoffisotope Deuterium und Tritium eingesetzt, die in der Fusionsreaktion zu Helium und einem Neutron reagieren. Da der magnetische Einschluss des Plasmas nicht perfekt ist, treffen auch Deuterium- und Tritiumionen auf die Berylliumwand auf und koennen sich dort anreichern. Das soll aufgrund der Radioaktivitaet von Tritium unbedingt vermieden werden. Daher wird zur Regenerierung der ersten Wand der Isotopenaustausch mit Deuterium untersucht. Wir

The Rocky Flats Environmental Technology Site beryllium characterization project

International Nuclear Information System (INIS)

Morrell, D.M.; Miller, J.R.; Allen, D.F.

1999-01-01

A site beryllium characterization project was completed at the Rocky Flats Environmental Technology Site (RFETS) in 1997. Information from historical reviews, previous sampling surveys, and a new sampling survey were used to establish a more comprehensive understanding of the locations and levels of beryllium contamination in 35 buildings. A feature of the sampling strategy was to test if process knowledge was a good predictor of where beryllium contamination could be found. Results revealed that this technique was effective at identifying where surface contamination levels might exceed the RFETS smear control level but that it was not effective in identifying where low concentrations of beryllium might be found

Phosphorus-containing azo compounds as analytical reagents for beryllium

International Nuclear Information System (INIS)

Lisenko, N.F.; Dolzhnikova, E.N.; Petrova, G.S.; Tsvetkov, E.N.; Vsesoyuznyj Nauchno-Issledovatel'skij Inst. Khimicheskikh Reaktivov i Osobo Chistykh Veshchestv, Moscow; AN SSSR, Moscow. Inst. Ehlementoorganicheskikh Soedinenij)

1979-01-01

The interaction of beryllium with six new azocompounds based on chromotropic or R-acids and o-aminophenyl-phenylphosphonic acids is studied. A sharp difference in the detection limit for beryllium by the two groups of compounds is found. Azoderivatives based on chromotropic acid are promising agent for beryllium due to sufficiently high selectivity. The introduction of the methyl-group into the o-position of the phosphorus-containing group improves the analytical properties of agents. Techniques are developed for the determination of beryllium in bronze, sewage water and in an artificial mixture using a sodium salt of 1.8-dioxi-2 [2' - (oxi- (o-methylphenyl)-phosphenyl)-phenilazo]-naphtalene-3.6-disulfoacid

Influence of physicochemical properties of beryllium particles on cultured cell toxicity

International Nuclear Information System (INIS)

Finch, G.L.; Brooks, A.L.; Hoover, M.D.; Cuddihy, R.G.

1988-01-01

The toxicity of beryllium oxide (BeO)), beryllium metal, and beryllium sulfate (BeSO 4 ) was studied in two cell lines, Chinese hamster ovary cells (CHO) and lung epithelial cells (LEC). Beryllium oxide particles were prepared at either 500 or 1000 deg. C, and two different particle sizes of beryllium metal were used. Following a 20-h exposure to beryllium compounds, cells were grown in culture to quantitate cloning ability relative to controls as a measure of cell killing, The LEC cultures were more sensitive to beryllium cytotoxicity than the CHO cells. When expressed on the basis of the mass of material added to the cultures, the order of toxicity was BeSO 4 ≥ 500 deg. C -BeO > 1000 deg. C -BeO > Be metal (small) Be metal (large). When cytotoxic effects were expressed on the basis of particulate surface rather than mass, the relative differences in toxicity between compounds was decreased. The order of toxicity was Be metal (small) ∼ Be metal (large) ∼ 500 deg. C-BeO ∼ 1000 deg. C-BeO. These data indicate that solubility influences beryllium toxicity to short-term cell cultures. (author)

Report on antimony-beryllium neutron sources examined after 7.45 effective full power years in the Dodewaard BWR

International Nuclear Information System (INIS)

Engen, W.R. van; Ainsworth, A.

1984-01-01

A source of neutrons is required to check instrumentation in the Dodewaard BWR after each shut-down. The longest shut-down is the annual refuelling shut-down when the reactor is routinely out for several weeks. Thus the isotopic neutron source needs to have a radioactive half-life which is sufficiently longer than several weeks. An appropriate source is antimony-124/ beryllium which has a half-life of 60.9 days. In this source 1.7 MeV gamma rays emitted by antimony-124 bombard the beryllium target material and produce neutrons. A report on the sources falls under the headings: design considerations (active insert; encapsulation (general considerations; chemical considerations; single or double encapsulation; radium-226 emanating source)); source production; post irradiation examination results (visual examination; eddy current tests); conclusions. (U.K.)

Validation of cleaning method for various parts fabricated at a Beryllium facility

Energy Technology Data Exchange (ETDEWEB)

Davis, Cynthia M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-12-15

This study evaluated and documented a cleaning process that is used to clean parts that are fabricated at a beryllium facility at Los Alamos National Laboratory. The purpose of evaluating this cleaning process was to validate and approve it for future use to assure beryllium surface levels are below the Department of Energy’s release limits without the need to sample all parts leaving the facility. Inhaling or coming in contact with beryllium can cause an immune response that can result in an individual becoming sensitized to beryllium, which can then lead to a disease of the lungs called chronic beryllium disease, and possibly lung cancer. Thirty aluminum and thirty stainless steel parts were fabricated on a lathe in the beryllium facility, as well as thirty-two beryllium parts, for the purpose of testing a parts cleaning method that involved the use of ultrasonic cleaners. A cleaning method was created, documented, validated, and approved, to reduce beryllium contamination.

The thorium fuel cycle in water-moderated reactor systems

International Nuclear Information System (INIS)

Critoph, E.

1977-01-01

Current interest in the thorium cycle, as an alternative to the uranium cycle, for water-moderated reactors is based on two attractive aspects of its use - the extension of uranium resources, and the related lower sensitivity of energy costs to uranium price. While most of the scientific basis required is already available, some engineering demonstrations are needed to provide better economic data for rational decisions. Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. There appear to be no major feasibility problems associated with the use of thorium, although development is required in the areas of fuel testing and fuel management. The use of thorium cycles implies recycling the fuel, and the major uncertainties are in the associated costs. Experience in the design and operation of fuel reprocessing and active-fabrication facilities is required to estimate costs to the accuracy needed for adequately defining the range of conditions economically favourable to thorium cycles. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An ''inventory'' of uranium of between 1 and 2Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium), is some two decades

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Compatibility behavior of beryllium with LiAlO2 and Li2ZrO3 ceramics, with 316L and 1.4914 steels in Sibelius

International Nuclear Information System (INIS)

Flament, T.; Roux, N.; Abassin, J.J.; Briec, M.; Cruz, D.; Schuster, I.

1991-01-01

The compatibility under irradiation of beryllium with Li 2 O, LiAlO 2 , Li 4 SiO 4 and Li 2 ZrO 3 ceramics and with 316L and 1.4914 steels was investigated in SIBELIUS. The irradiation was performed in the SILOE reactor at 550 deg C for 1690 hours in He + 0.1%H 2 purge pas. Examinations of the LiAlO 2 /Be and Li 2 ZrO 3 /Be couples show a weak oxidation of beryllium and the presence of cavities near the interface with ceramics. Examinations of the 316L/Be and 1.4914/Be couples show the formation of an oxide layer on all beryllium and steel surfaces suggesting that corrosion arises from a species (most likely T 2 O and/or H 2 O) present in the environmental atmosphere. Post-irradiation annealing tests of beryllium indicate that the major part of helium is released during irradiation whereas the major part of tritium is released above 700 deg C

Extraction of beryllium sulfate by a long chain amine; Extraction du sulfate de beryllium par une amine a longue chaine

Energy Technology Data Exchange (ETDEWEB)

Etaix, E.S. [Commissariat a l' Energie Atomique, Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

1968-06-01

The extraction of sulfuric acid in aqueous solution by a primary amine in benzene solution, 3-9 (diethyl) - 6-amino tri-decane (D.E.T. ) - i.e., with American nomenclature 1-3 (ethyl-pentyl) - 4-ethyl-octyl amine (E.P.O.) - has made it possible to calculate the formation constants of alkyl-ammonium sulfate and acid sulfate. The formula of the beryllium and alkyl-ammonium sulfate complex formed in benzene has next been determined, for various initial acidity of the aqueous solution. Lastly, evidence has been given of negatively charged complexes of beryllium and sulfate in aqueous solution, through the dependence of the aqueous sulfate ions concentration upon beryllium extraction. The formation constant of these anionic complexes has been evaluated. (author) [French] L'etude de l'extraction de l'acide sulfurique en solution aqueuse par une amine primaire en solution dans le benzene, le diethyl-3,9 amino-6 tridecane (D.E.T.) - autre nom americain 1-3 (ethylpentyl) - 4-ethyloctylamine (E.P.O.) a permis de calculer les constantes de formation du sulfate et de l'hydrogenosulfate d'alkyl-ammonium. La formule du complexe de sulfate de beryllium et d'alkyl-ammonium forme en solution benzenique a ete ensuite determinee pour diverses acidites initiales de la solution aqueuse. Enfin, l'influence de la concentration des ions sulfate de la phase aqueuse sur l'extraction du beryllium a mis en evidence la formation en solution aqueuse de complexes anioniques de sulfate et de beryllium dont la constante de formation a ete evaluee. (auteur)

Extraction of beryllium sulfate by a long chain amine; Extraction du sulfate de beryllium par une amine a longue chaine

Energy Technology Data Exchange (ETDEWEB)

Etaix, E S [Commissariat a l' Energie Atomique, Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

1968-06-01

The extraction of sulfuric acid in aqueous solution by a primary amine in benzene solution, 3-9 (diethyl) - 6-amino tri-decane (D.E.T. ) - i.e., with American nomenclature 1-3 (ethyl-pentyl) - 4-ethyl-octyl amine (E.P.O.) - has made it possible to calculate the formation constants of alkyl-ammonium sulfate and acid sulfate. The formula of the beryllium and alkyl-ammonium sulfate complex formed in benzene has next been determined, for various initial acidity of the aqueous solution. Lastly, evidence has been given of negatively charged complexes of beryllium and sulfate in aqueous solution, through the dependence of the aqueous sulfate ions concentration upon beryllium extraction. The formation constant of these anionic complexes has been evaluated. (author) [French] L'etude de l'extraction de l'acide sulfurique en solution aqueuse par une amine primaire en solution dans le benzene, le diethyl-3,9 amino-6 tridecane (D.E.T.) - autre nom americain 1-3 (ethylpentyl) - 4-ethyloctylamine (E.P.O.) a permis de calculer les constantes de formation du sulfate et de l'hydrogenosulfate d'alkyl-ammonium. La formule du complexe de sulfate de beryllium et d'alkyl-ammonium forme en solution benzenique a ete ensuite determinee pour diverses acidites initiales de la solution aqueuse. Enfin, l'influence de la concentration des ions sulfate de la phase aqueuse sur l'extraction du beryllium a mis en evidence la formation en solution aqueuse de complexes anioniques de sulfate et de beryllium dont la constante de formation a ete evaluee. (auteur)

Preparation of copper-beryllium alloys from Indian beryl

International Nuclear Information System (INIS)

Paul, C.M.; Sharma, B.P.; Subba Rao, K.S.; Rajadhyaksha, M.G.; Sundaram, C.V.

1975-01-01

The report presents the results of laboratory scale investigations on the preparation of copper-beryllium and aluminium-beryllium master alloys starting from Indian beryl and adopting the fluoride process. The flow-sheet involves : (1) conversion of the Be-values in beryl into water soluble sodium beryllium fluoride (2) preparation of beryllium hydroxide by alkali treatment of aqueous Na 2 BeF 4 (3) conversion of Be(OH) 2 to (NH 4 ) 2 BeF 4 by treatment with NH 4 HF 2 (4) thermal decomposition of (NH 4 ) 2 BeF 4 to BeF 2 and (5) magnesium reduction of BeF 2 (with the addition of copper/aluminium) to obtain beryllium alloys. The method has been successfully employed for the preparation of Cu-Be master alloys containing about 8% Be and free of Mg on a 200 gm scale. An overall Be-recovery of about 80% has been achieved. Al-8% Be master alloys have also been prepared by this method. Toxicity and health hazards associated with Be are discussed and the steps taken to ensure safe handling of Be are described. (author)

Erosion of beryllium under ITER - Relevant transient plasma loads

Science.gov (United States)

Kupriyanov, I. B.; Nikolaev, G. N.; Kurbatova, L. A.; Porezanov, N. P.; Podkovyrov, V. L.; Muzichenko, A. D.; Zhitlukhin, A. M.; Gervash, A. A.; Safronov, V. M.

2015-08-01

Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2-0.5 MJ/m2 at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology.

Reflector-moderated critical assemblies

International Nuclear Information System (INIS)

Paxton, H.C.; Jarvis, G.A.; Byers, C.C.

1975-07-01

Experiments with reflector-moderated critical assemblies were part of the Rover Program at the Los Alamos Scientific Laboratory (LASL). These assemblies were characterized by thick D 2 O or beryllium reflectors surrounding large cavities that contained highly enriched uranium at low average densities. Because interest in this type of system has been revived by LASL Plasma Cavity Assembly studies, more detailed descriptions of the early assemblies than had been available in the unclassified literature are provided. (U.S.)

Research on plasma core reactors

International Nuclear Information System (INIS)

Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

1977-01-01

Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation

Decontamination and decommissioning of the Organic Moderated Reactor Experiment facility (OMRE)

International Nuclear Information System (INIS)

Hine, R.E.

1980-09-01

This report describes the decontamination and decommissioning (D and D) of the Organic Moderated Reactor Experiment (OMRE) facility performed from October 1977 through September 1979. This D and D project included removal of all the facilities and as much contaminated soil and rock as practical. Removal of the reactor pressure vessel was an unusually difficult problem, and an extraordinary, unexpected amount of activated rock and soil was removed. After removal of all significantly contaminated material, the site consisted of a 20-ft deep excavation surrounded by backfill material. Before this excavation was backfilled, it and the backfill material were radiologically surveyed and detailed records made of these surveys. After the excavation was backfilled and graded, the site surface was surveyed again and found to be essentially uncontaminated

Status of beryllium development for fusion applications

International Nuclear Information System (INIS)

Billone, M.C.; Donne, M.D.; Macaulay-Newcombe, R.G.

1994-05-01

Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma facing component of first wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold isostatic pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation

Critical parameters controlling irradiation swelling in beryllium

International Nuclear Information System (INIS)

Dubinko, V.I.

1995-01-01

Radiation effects in beryllium can hardly be explained within a framework of the conventional theory based on the bias concept due to elastic interaction difference (EID) between vacancies and self-interstitial atoms (SIAs) since beryllium belongs to hexagonal close-packed metals where diffusion has been shown to be anisotropic. Diffusional anisotropy difference (DAD) between point defects changes the cavity bias for their absorption and leads to dependence of the dislocation bias on the distribution of dislocations over crystallographic directions. On the other hand, the elastic interaction between point defects and cavities gives rise to the size and gas pressure dependencies of the cavity bias, resulting in new critical quantities for bubble-void transition effects at low temperature irradiation. In the present paper, we develop the concept of the critical parameters controlling irradiation swelling with account of both DAD and EID, and take care of thermal effects as well since they are of major importance for beryllium which has an anomalously low self-diffusion activation energy. Experimental data on beryllium swelling are analyzed on the basis of the present theory. (orig.)

Beryllium. Health hazards and their control. Pt. 2

International Nuclear Information System (INIS)

Lires, O.A.; Delfino, C.A.; Botbol, J.

1991-01-01

In this work (continuation of 'Beryllium' series) health hazards, toxic effects, limits of permissible atmospheric contamination and safe exposure to beryllium are described. Guidelines to the design, control operations and hygienic precautions of the working facilities are given. (Author) [es

Status of the beryllium replacement project

Energy Technology Data Exchange (ETDEWEB)

Dimayuga, I. [Atomic Energy of Canada, Chalk River, Ontario (Canada); Corcoran, E. [Royal Military College of Canada, Kingston, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Harmsen, J. [Cameco Fuel Manufacturing Inc., Port Hope, Ontario (Canada); Lu, E. [Bruce Power, Tiverton, Ontario (Canada); Onderwater, T.; Palleck, S. [General Electric- Hitachi, Wilmington, North Carolina (United States); Pant, A. [Cameco Fuel Manufacturing Inc., Port Hope, Ontario (Canada)

2013-07-01

Currently, beryllium (Be) is used as the filler metal for brazing appendages on the sheaths of CANDU® fuel elements. Because of its toxicity, occupational exposure limits for Be are being reduced to very low levels, resulting in significant challenges to CANDU® fuel fabricators. The CANDU® Owners Group (COG) initiated a test program to identify a filler material to replace Be and confirm that the brazed joints meet the established technical requirements for CANDU® fuel. Together with eliminating health risks associated with the use of Be, the industry needs to be assured that continuation of fuel supply remains unaffected and that fuel fabrication processes continue to comply with health and safety standards. A literature survey of studies on brazing and joining of Zircaloy identified potential filler materials that can meet or exceed existing design requirements of the brazed joint, including the required mechanical, microstructural, corrosion resistance, and irradiation properties equivalent to those obtained with Be as braze material. Candidate materials were evaluated against several criteria, including manufacturability, melting point, wettability, mechanical properties, corrosion resistance, effect on neutron economy, potential activation products, and interaction with fuel channels and other related disciplines. This exercise resulted in a list of promising candidate materials that were recommended for the first phase of testing. These materials include stainless steel (304 or 316), Al-Si, Ni-P, and Zr-Mn alloys. To allow a CANDU® utility have sufficient confidence in considering implementation of a different braze filler material, a Be Replacement Test Program, involving out-reactor and in-reactor tests, is being undertaken as a collaborative endeavour by the Canadian nuclear industry. The out-reactor tests consist of: a constructability assessment to determine the material’s suitability with current fuel manufacturing methods; evaluation of

Beryllium phonon spectrum from cold neutron measurements

International Nuclear Information System (INIS)

Bulat, I.A.

1979-01-01

The inelastic coherent scattering of neutrons with the initial energy E 0 =4.65 MeV on the spectrometer according to the time of flight is studied in polycrystalline beryllium. The measurements are made for the scattering angles THETA=15, 30, 45, 60, 75 and 90 deg at 293 K. The phonon spectrum of beryllium, i-e. g(w) is reestablished from the experimental data. The data obtained are compared with the data of model calculations. It is pointed out that the phonon spectrum of beryllium has a bit excessive state density in the energy range from 10 to 30 MeV. It is caused by the insufficient statistical accuracy of the experiment at low energy transfer

ICT diagnostic method of beryllium welding quality

International Nuclear Information System (INIS)

Sun Lingxia; Wei Kentang; Ye Yunchang

2002-01-01

To avoid the interference of high density material for the quality assay of beryllium welding line, a slice by slice scanning method was proposed based upon the research results of the Industrial Computerized Tomography (ICT) diagnostics for weld penetration, weld width, off-centered deviation and weld defects of beryllium-ring welding seam with high density material inside

Tensile and fracture toughness test results of neutron irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Chaouadi, R.; Moons, F.; Puzzolante, J.L. [Centre d`Etude de l`Energie Nucleaire, Mol (Belgium)

1998-01-01

Tensile and fracture toughness test results of four Beryllium grades are reported here. The flow and fracture properties are investigated by using small size tensile and round compact tension specimens. Irradiation was performed at the BR2 material testing reactor which allows various temperature and irradiation conditions. The fast neutron fluence (>1 MeV) ranges between 0.65 and 2.45 10{sup 21} n/cm{sup 2}. In the meantime, un-irradiated specimens were aged at the irradiation temperatures to separate if any the effect of temperature from irradiation damage. Test results are analyzed and discussed, in particular in terms of the effects of material grade, test temperature, thermal ageing and neutron irradiation. (author)

Dosage of boron traces in graphite, uranium and beryllium oxide

International Nuclear Information System (INIS)

Coursier, J.; Hure, J.; Platzer, R.

1955-01-01

The problem of the dosage of the boron in the materials serving to the construction of nuclear reactors arises of the following way: to determine to about 0,1 ppm close to the quantities of boron of the order of tenth ppm. We have chosen the colorimetric analysis with curcumin as method of dosage. To reach the indicated contents, it is necessary to do a previous separation of the boron and the materials of basis, either by extraction of tetraphenylarsonium fluoborate in the case of the boron dosage in uranium and the beryllium oxide, either by the use of a cations exchanger resin of in the case of graphite. (M.B.) [fr

Application of noise analysis technique for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors

International Nuclear Information System (INIS)

Shieh, D.J.; Upadhyaya, B.R.; Sweeney, F.J.

1987-01-01

A new technique, based on the noise analysis of neutron detector and core-exit coolant temperature signals, is developed for monitoring the moderator temperature coefficient of reactivity in pressurized water reactors (PWRs). A detailed multinodal model is developed and evaluated for the reactor core subsystem of the loss-of-fluid test (LOFT) reactor. This model is used to study the effect of changing the sign of the moderator temperature coefficient of reactivity on the low-frequency phase angle relationship between the neutron detector and the core-exit temperature noise signals. Results show that the phase angle near zero frequency approaches - 180 deg for negative coefficients and 0 deg for positive coefficients when the perturbation source for the noise signals is core coolant flow, inlet coolant temperature, or random heat transfer

Mixed core management: Use of 93% and 72% enriched uranium in the BR2 reactor

International Nuclear Information System (INIS)

Ponsard, B.

2000-01-01

The BR2 reactor, put into operation in 1963 and refurbished from July 1995 till April 1997, is a 100 MW high-flux Materials Testing Reactor, using 93% 235 U enriched uranium as standard fuel, light water as coolant and beryllium as moderator. The present operating regime consists of five irradiation cycles per year at an operating power between 50 and 70 MW; each cycle is characterized by 21 days operation. In the framework of a 'qualification programme', six 72% 235 U fuel elements fabricated with uranium recovered from the reprocessing of BR2 spent fuel at UKAEA-Dounreay have been successfully irradiated in the period 1994-1995 reaching a maximum mean burnup of 48% without the release of fission products. Since 1998, this type of fuel element is irradiated routinely together with standard 93% 235 U fuel elements in order to optimize the utilization of the available HEU inventory. The purpose of this paper is to present the strategy developed in order to optimize the mixed core management of the BR2 reactor. (author)

5. IEA International workshop on beryllium technology for fusion. Book of abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be{sub 12}Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented.

5. IEA International workshop on beryllium technology for fusion. Book of abstracts

International Nuclear Information System (INIS)

2001-01-01

The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be 12 Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented [ru

A plan of reactor physics experiments for reduced-moderation water reactors with MOX fuel in TCA

International Nuclear Information System (INIS)

Shimada, Shoichiro; Akie, Hiroshi; Suzaki, Takenori; Okubo, Tutomu; Usui, Shuji; Shirakawa, Toshihisa; Iwamura, Takamiti; Kugo, Teruhiko; Ishikawa, Nobuyuki

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors which aim at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. Critical Experiments performed so far in Eualope and Japan were reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of the equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX fuel rods used in the experiments are obtained by calculations and the modification of the equipment for the experiments are shown. (author)

Tritium behavior in ITER beryllium

International Nuclear Information System (INIS)

Longhurst, G.R.

1990-10-01

The beryllium neutron multiplier in the ITER breeding blanket will generate tritium through transmutations. That tritium constitutes a safety hazard. Experiments evaluating tritium storage and release mechanisms have shown that most of the tritium comes out in a burst during thermal ramping. A small fraction of retained tritium is released by thermally activated processes. Analysis of recent experimental data shows that most of the tritium resides in helium bubbles. That tritium is released when the bubbles undergo swelling sufficient to develop porosity that connects with the surface. That appears to occur when swelling reaches about 10--15%. Other tritium appears to be stored chemically at oxide inclusions, probably as Be(OT) 2 . That component is released by thermal activation. There is considerable variation in published values for tritium diffusion through the beryllium and solubility in it. Data from experiments using highly irradiated beryllium from the Idaho National Engineering Laboratory showed diffusivity generally in line with the most commonly accepted values for fully dense material. Lower density material, planned for use in the ITER blanket may have very short diffusion times because of the open structure. The beryllium multiplier of the ITER breeding blanket was analyzed for tritium release characteristics using temperature and helium production figures at the midplane generated in support of the ITER Summer Workshop, 1990 in Garching. Ordinary operation, either in Physics or Technology phases, should not result in the release of tritium trapped in the helium bubbles. Temperature excursions above 600 degree C result in large-scale release of that tritium. 29 refs., 10 figs., 3 tabs

Nuclear reactor, reactor core thereof, and device for constituting the reactor

International Nuclear Information System (INIS)

Takiyama, Masashi.

1994-01-01

A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

Study of the microstructure of neutron irradiated beryllium for the validation of the ANFIBE code

International Nuclear Information System (INIS)

Rabaglino, E.; Ferrero, C.; Reimann, J.; Ronchi, C.; Schulenberg, T.

2002-01-01

The behaviour of beryllium under fast neutron irradiation is a key issue of the helium cooled pebble bed tritium breeding blanket, due to the production of large quantities of helium and of a non-negligible amount of tritium. To optimise the design, a reliable prediction of swelling due to helium bubbles and of tritium inventory during normal and off-normal operation of a fusion power reactor is needed. The ANFIBE code (ANalysis of Fusion Irradiated BEryllium) is being developed to meet this need. The code has to be applied in a range of irradiation conditions where no experimental data are available, therefore a detailed gas kinetics model, and a specific and particularly careful validation strategy are needed. The validation procedure of the first version of the code was based on macroscopic data of swelling and tritium release. This approach is, however, incomplete, since a verification of the microscopic behaviour of the gas in the metal is necessary to obtain a reliable description of swelling. This paper discusses a general strategy for a thorough validation of the gas kinetics models in ANFIBE. The microstructure characterisation of weakly irradiated beryllium pebbles, with different visual examination techniques, is then presented as an example of the application of this strategy. In particular, the advantage of developing 3D techniques, such as X-ray microtomography, is demonstrated

Novel plasma source for safe beryllium spectral line studies in the presence of beryllium dust

Science.gov (United States)

Stankov, B. D.; Vinić, M.; Gavrilović Božović, M. R.; Ivković, M.

2018-05-01

Plasma source for beryllium spectral line studies in the presence of beryllium dust particles was realised. The guideline during construction was to prevent exposure to formed dust, considering the toxicity of beryllium. Plasma source characterization through determination of optimal working conditions is described. The necessary conditions for Be spectral line appearance and optimal conditions for line shape measurements are found. It is proven experimentally that under these conditions dust appears coincidently with the second current maximum. The electron density measured after discharge current maximum is determined from the peak separation of the hydrogen Balmer beta spectral line, and the electron temperature is determined from the ratios of the relative intensities of Be spectral lines emitted from successive ionized stages of atoms. Maximum values of electron density and temperature are measured to be 9.3 × 1022 m-3 and 16 800 K, respectively. Construction details and testing of the BeO discharge tube in comparison with SiO2 and Al2O3 discharge tubes are also presented in this paper.

Measurement of neutron disadvantage factor for fuel and moderator in the square reactor cell

International Nuclear Information System (INIS)

Bosevski, T.; Spiric, V.

1964-01-01

Full text: Heterogeneous diffusion treatment for flux distribution was used to define the direction of measurements for obtaining mean neutron flux in the moderator of the reactor cell by single integration. Factor Q for the fuel was determined by using experimental flux distribution in the cell moderator and calculated values for the function X (x;y). Experimental and calculated results are shown as a diagram. All the calculations were done on the ZUSE-Z-23 computer

Hydrodynamic instabilities in beryllium targets for the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Yi, S. A., E-mail: austinyi@lanl.gov; Simakov, A. N.; Wilson, D. C.; Olson, R. E.; Kline, J. L.; Batha, S. H. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States); Clark, D. S.; Hammel, B. A.; Milovich, J. L.; Salmonson, J. D.; Kozioziemski, B. J. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551 (United States)

2014-09-15

Beryllium ablators offer higher ablation velocity, rate, and pressure than their carbon-based counterparts, with the potential to increase the probability of achieving ignition at the National Ignition Facility (NIF) [E. I. Moses et al., Phys. Plasmas 16, 041006 (2009)]. We present here a detailed hydrodynamic stability analysis of low (NIF Revision 6.1) and high adiabat NIF beryllium target designs. Our targets are optimized to fully utilize the advantages of beryllium in order to suppress the growth of hydrodynamic instabilities. This results in an implosion that resists breakup of the capsule, and simultaneously minimizes the amount of ablator material mixed into the fuel. We quantify the improvement in stability of beryllium targets relative to plastic ones, and show that a low adiabat beryllium capsule can be at least as stable at the ablation front as a high adiabat plastic target.

Summary report of the 7th reduced-moderation water reactor workshop

International Nuclear Information System (INIS)

Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

2005-08-01

As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (author)

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Structure/property relationships in multipass GMA welding of beryllium.

Energy Technology Data Exchange (ETDEWEB)

Hochanadel, P. W. (Patrick W.); Hults, W. L. (William L.); Thoma, D. J. (Dan J.); Dave, V. R. (Vivek R.); Kelly, A. M. (Anna Marie); Pappin, P. A. (Pallas A.); Cola, M. J. (Mark J.); Burgardt, P. (Paul)

2001-01-01

Beryllium is an interesting metal that has a strength to weight ratio six times that of steel. Because of its unique mechanical properties, beryllium is used in aerospace applications such as satellites. In addition, beryllium is also used in x-ray windows because it is nearly transparent to x-rays. Joining of beryllium has been studied for decades (Ref.l). Typically joining processes include braze-welding (either with gas tungsten arc or gas metal arc), soldering, brazing, and electron beam welding. Cracking which resulted from electron beam welding was recently studied to provide structure/property relationships in autogenous welds (Ref. 2). Braze-welding utilizes a welding arc to melt filler, and only a small amount of base metal is melted and incorporated into the weld pool. Very little has been done to characterize the braze-weld in terms of the structure/property relationships, especially with reference to multipass welding. Thus, this investigation was undertaken to evaluate the effects of multiple passes on microstructure, weld metal composition, and resulting material properties for beryllium welded with aluminum-silicon filler metal.

The results of medical surveillance of beryllium production personnel