Reactor production of 252Cf and transcurium isotopes

International Nuclear Information System (INIS)

Alexander, C.W.; Halperin, J.; Walker, R.L.; Bigelow, J.E.

1990-01-01

Berkelium, californium, einsteinium, and fermium are currently produced in the High Flux Isotope Reactor (HFIR) and recovered in the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). All the isotopes are used for research. In addition, 252 Cf, 253 Es, and 255 Fm have been considered or are used for industrial or medical applications. ORNL is the sole producer of these transcurium isotopes in the western world. A wide range of actinide samples were irradiated in special test assemblies at the Fast Flux Test Facility (FFTF) at Hanford, Washington. The purpose of the experiments was to evaluate the usefulness of the two-group flux model for transmutations in the special assemblies with an eventual goal of determining the feasibility of producing macro amounts of transcurium isotopes in the FFTF. Preliminary results from the production of 254g Es from 252 Cf will be discussed. 14 refs., 5 tabs

Safety issues at the defense production reactors

International Nuclear Information System (INIS)

1987-01-01

The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors endash the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. The report is necessarily based on a limited review of the defense production reactors. It does not address whether any of the reactors are ''safe,'' because such an analysis would involve a determination of acceptable risk endash a matter of obvious importance, but one that was beyond the purview of the committee. It also does not address whether the safety of the production reactors is comparable to that of commercial nuclear power stations, because even this narrower question extended beyond the charge to the committee and would have involved detailed analyses that the committee could not undertake

Hydrogen production in fusion reactors

Science.gov (United States)

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

1993-11-01

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

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Malkoske, G.R.; Norton, J.L.; Slack, J.

2002-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

Cobalt-60 production in CANDU power reactors

International Nuclear Information System (INIS)

Slack, J.; Norton, J.L.; Malkoske, G.R.

2003-01-01

MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments. The major application is in the health care industry where irradiators are used to sterilize single use medical products. These irradiators are designed and built by MDS Nordion and are used by manufacturers of surgical kits, gloves, gowns, drapes and other medical products. The irradiator is a large shielded room with a storage pool for the cobalt-60 sources. The medical products are circulated through the shielded room and exposed to the cobalt-60 sources. This treatment sterilizes the medical products which can then be shipped to hospitals for immediate use. Other applications for this irradiation technology include sanitisation of cosmetics, microbial reduction of pharmaceutical raw materials and food irradiation. The cobalt-60 sources are manufactured by MDS Nordion in their Cobalt Operations Facility in Kanata. More than 75,000 cobalt-60 sources for use in irradiators have been manufactured by MDS Nordion. The cobalt-60 sources are double encapsulated in stainless steel capsules, seal welded and helium leak tested. Each source may contain up to 14,000 curies. These sources are shipped to over 170 industrial irradiators around the world. This paper will focus on the MDS Nordion proprietary technology used to produce the cobalt-60 isotope in CANDU reactors. Almost 55 years ago MDS Nordion and Atomic Energy of Canada developed the process for manufacturing cobalt-60 at the Chalk River Labs, in Ontario, Canada. A cobalt-59 target was introduced into a research reactor where the cobalt-59 atom absorbed one neutron to become cobalt-60. Once the cobalt-60 material was removed from the research reactor it was encapsulated in stainless steel and seal welded using a Tungsten Inert Gas weld. The first cobalt-60 sources manufactured using material from the Chalk River Labs were used in cancer

High temperature reactor for the production of low temperature heat

International Nuclear Information System (INIS)

Muehlensiep, J.

1986-12-01

In this report the conditions of nuclear working reactors for district heating are described for the use in suburban areas. The design of a HTR is analysed under the point of view of safety and costs for the components and for the arrangement possibilities. The size of system is chosen by analysing important parameters for construction. The layout is determined by the retention of fission products in the coated particles of the fuel under conditions of hypothetical accidents. Based on stated data a HTR reactor for district heating will be designed. The speciality is a square shaped core which has the advantage to conduct the afterheat fastly to the outside of the pressure vessel in case of hypothetical accidents. Caused by the shape of the core the heat exchangers may be installed next to the core, the shutdown rods are maintained into reflector borings where they have a high efficiency. The whole primary circuit is surrounded by the reactor pressure vessel and is adjusted in an underground concrete cell. (orig./GL) [de

Hydrogen production in fusion reactors

International Nuclear Information System (INIS)

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

1993-11-01

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

Hydrogen production in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

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

1993-11-01

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

Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

International Nuclear Information System (INIS)

1986-11-01

This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

Sodium fast reactors with closed fuel cycle

CERN Document Server

Raj, Baldev; Vasudeva Rao, PR 0

2015-01-01

Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

Product competitiveness analysis for e-commerce platform of special agricultural products

Science.gov (United States)

Wan, Fucheng; Ma, Ning; Yang, Dongwei; Xiong, Zhangyuan

2017-09-01

On the basis of analyzing the influence factors of the product competitiveness of the e-commerce platform of the special agricultural products and the characteristics of the analytical methods for the competitiveness of the special agricultural products, the price, the sales volume, the postage included service, the store reputation, the popularity, etc. were selected in this paper as the dimensionality for analyzing the competitiveness of the agricultural products, and the principal component factor analysis was taken as the competitiveness analysis method. Specifically, the web crawler was adopted to capture the information of various special agricultural products in the e-commerce platform ---- chi.taobao.com. Then, the original data captured thereby were preprocessed and MYSQL database was adopted to establish the information library for the special agricultural products. Then, the principal component factor analysis method was adopted to establish the analysis model for the competitiveness of the special agricultural products, and SPSS was adopted in the principal component factor analysis process to obtain the competitiveness evaluation factor system (support degree factor, price factor, service factor and evaluation factor) of the special agricultural products. Then, the linear regression method was adopted to establish the competitiveness index equation of the special agricultural products for estimating the competitiveness of the special agricultural products.

Reactor physics special problem in 11. ENFIR

International Nuclear Information System (INIS)

Leszczynski, Francisco

1997-01-01

In this report, the computation method and the results of the work performed of the special topic on reactor physics proposed for the 11. ENFIR is presented. MCNP 4.2 has been adopted as the only code to perform the calculations. The full core of the IPEN-MB-1 critical unit has been modelled without important approximations. The specifications given by the Organizer Commission of the Special Topic were followed. The nuclear libraries adopted were those included on the MCNPDAT package, mainly from ENDF/B-V, except indium data, not included in this package. For indium, data obtained from LANL, based on ENDF/B-VI were used. The results are: critical position of the control banks assuming simultaneous movement: percent of extraction: (49±1)% ; excess of reactivity of the core: ρ =( 3590 ±50)pcm ; total reactivity of the one control rod bank: ρ= (4000±50) pcm. The reactivity curve of the control rods is included also. (author)

Innovative self-powered submersible microbial electrolysis cell (SMEC) for biohydrogen production from anaerobic reactors

DEFF Research Database (Denmark)

Zhang, Yifeng; Angelidaki, Irini

2012-01-01

A self-powered submersible microbial electrolysis cell (SMEC), in which a specially designed anode chamber and external electricity supply were not needed, was developed for in situ biohydrogen production from anaerobic reactors. In batch experiments, the hydrogen production rate reached 17.8 m...... improvement of voltage output and reduction of electron losses were essential for efficient hydrogen generation. In addition, alternate exchanging the electricity-assisting and hydrogen-producing function between the two cell units of the SMEC was found to be an effective approach to inhibit methanogens...

Pebble Bed Reactor Dust Production Model

Energy Technology Data Exchange (ETDEWEB)

Abderrafi M. Ougouag; Joshua J. Cogliati

2008-09-01

The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors’ PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production.

Pebble Bed Reactor Dust Production Model

International Nuclear Information System (INIS)

Abderrafi M. Ougouag; Joshua J. Cogliati

2008-01-01

The operation of pebble bed reactors, including fuel circulation, can generate graphite dust, which in turn could be a concern for internal components; and to the near field in the remote event of a break in the coolant circuits. The design of the reactor system must, therefore, take the dust into account and the operation must include contingencies for dust removal and for mitigation of potential releases. Such planning requires a proper assessment of the dust inventory. This paper presents a predictive model of dust generation in an operating pebble bed with recirculating fuel. In this preliminary work the production model is based on the use of the assumption of proportionality between the dust production and the normal force and distance traveled. The model developed in this work uses the slip distances and the inter-pebble forces computed by the authors PEBBLES. The code, based on the discrete element method, simulates the relevant static and kinetic friction interactions between the pebbles as well as the recirculation of the pebbles through the reactor vessel. The interaction between pebbles and walls of the reactor vat is treated using the same approach. The amount of dust produced is proportional to the wear coefficient for adhesive wear (taken from literature) and to the slip volume, the product of the contact area and the slip distance. The paper will compare the predicted volume with the measured production rates. The simulation tallies the dust production based on the location of creation. Two peak production zones from intra pebble forces are predicted within the bed. The first zone is located near the pebble inlet chute due to the speed of the dropping pebbles. The second peak zone occurs lower in the reactor with increased pebble contact force due to the weight of supported pebbles. This paper presents the first use of a Discrete Element Method simulation of pebble bed dust production

Possibility of using gamma radiation from HTR reactors for the processing of food and medical products

International Nuclear Information System (INIS)

Pahladsingh, R.R.

2004-01-01

During the fission process in most of the presently operating nuclear reactors nuclear energy is converted into thermal energy and transferred to common steam cycles for power generation. As part of the fission process also α, β and neutrons particles are released from the nucleus; the release of gamma-rays is also a part of the fission process. In present nuclear reactors α, β, neutrons particles and particularly gamma-rays are not gainfully used as a result of the reactor design and of the containment. These plants are built as required by regulations and international standards for safety. The inherently safe HTR reactor, by its physics and design, does not need a special reinforced containment and it is worth looking into the possibilities of this design feature to use the by-products, such as Gamma-rays, from nuclear fission. In the HTR Pebble Bed Reactors the α, and β particles will remain in the kernels of the pebbles. This means that only the neutron particles and gamma-rays will be available outside the reactor pressure vessel. In this report a proposal is presented to use the gamma-rays of the HTR reactor for irradiation of food and agricultural produce. For neutron shielding a reflector is placed inside the reactor while outside the reactor neutron- and thermal-shielding will be accomplished with water. The high energy gamma-rays will pass through the water-shield and could be harnessed for radiation processing of food and medical products. (author)

The behavior of fission products during nuclear rocket reactor tests

International Nuclear Information System (INIS)

Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

1991-01-01

Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

Special servicing equipment for reactor pressurized vessel stud hole and stud accessories

International Nuclear Information System (INIS)

Li Jianglian

1999-01-01

The author briefly introduces the design and manufacture of nuclear island special servicing equipment of Nuclear Power Institute of China. Maintenance process of reactor pressurized vessel (RPV) stud hold and stud accessories the special servicing equipment include RPV flange dummy, closed-circuit television (CCTV) inspection equipment, RPV stud hole expandable comb, RPV stud hole polisher, RPV stud hold thread lubricating equipment, RPV stud hole thread miller and RPV stud hole camera. It is presented how eight kinds of special servicing equipment perform the maintenance process concerning their function, structure, and characteristics, their practical use on site is also introduced

Reactor calculations in aid of isotope production at SAFARI-1

International Nuclear Information System (INIS)

Ball, G.

2003-01-01

Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

Feasibility study of a magnetic fusion production reactor

Science.gov (United States)

Moir, R. W.

1986-12-01

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

Special-purpose materials for magnetically confined fusion reactors. Third annual progress report

International Nuclear Information System (INIS)

1981-11-01

The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. It is recognized that there will be numerous materials problems that will arise during the design and construction of large magnetic-fusion energy devices such as the Engineering Test Facility (ETF) and Demonstration Reactor (DEMO). Most of these problems will be specific to a particular design or project and are the responsibility of the project, not the Materials and Radiation Effects Branch. Consequently, the Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits

Nuclear power reactors: reactor safety and military and civil defence

International Nuclear Information System (INIS)

Hvinden, T.

1976-01-01

The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

Radioactive effluent sources and special system of channels on the RA Reactor at the Boris Kidric Institute

International Nuclear Information System (INIS)

Bojovic, P.; Gacinovic, O.; Milosevic, M.

1964-10-01

The paper describes the place of origin, composition and activity of radioactive effluents appearing in some reactor systems and special channels for carrying these effluents to disposal basins located outside the reactor building (author)

Target-fueled nuclear reactor for medical isotope production

Science.gov (United States)

Coats, Richard L.; Parma, Edward J.

2017-06-27

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

Method of producing gaseous products using a downflow reactor

Science.gov (United States)

Cortright, Randy D; Rozmiarek, Robert T; Hornemann, Charles C

2014-09-16

Reactor systems and methods are provided for the catalytic conversion of liquid feedstocks to synthesis gases and other noncondensable gaseous products. The reactor systems include a heat exchange reactor configured to allow the liquid feedstock and gas product to flow concurrently in a downflow direction. The reactor systems and methods are particularly useful for producing hydrogen and light hydrocarbons from biomass-derived oxygenated hydrocarbons using aqueous phase reforming. The generated gases may find used as a fuel source for energy generation via PEM fuel cells, solid-oxide fuel cells, internal combustion engines, or gas turbine gensets, or used in other chemical processes to produce additional products. The gaseous products may also be collected for later use or distribution.

Activation product transport in fusion reactors

International Nuclear Information System (INIS)

Klein, A.C.

1983-01-01

Activated corrosion and neutron sputtering products will enter the coolant and/or tritium breeding material of fusion reactor power plants and experiments and cause personnel access problems. Radiation levels around plant components due to these products will cause difficulties with maintenance and repair operations throughout the plant. Similar problems are experienced around fission reactor systems. The determination of the transport of radioactive corrosion and neutron sputtering products through the system is achieved using the computer code RAPTOR. This code calculates the mass transfer of a number of activation products based on the corrosion and sputtering rates through the system, the deposition and release characteristics of various plant components, the neturon flux spectrum, as well as other plant parameters. RAPTOR assembles a system of first order linear differential equations into a matrix equation based upon the reactor system parameters. Included in the transfer matrix are the deposition and erosion coefficients, and the decay and activation data for the various plant nodes and radioactive isotopes. A source vector supplies the corrosion and neutron sputtering source rates. This matrix equation is then solved using a matrix operator technique to give the specific activity distribution of each radioactive species throughout the plant. Once the amount of mass transfer is determined, the photon transport due to the radioactive corrosion and sputtering product sources can be evaluated, and dose rates around the plant components of interest as a function of time can be determined. This method has been used to estimate the radiation hazards around a number of fusion reactor system designs

Measurement of tritium production rate distribution for a fusion-fission hybrid conceptual reactor

International Nuclear Information System (INIS)

Wang Xinhua; Guo Haiping; Mou Yunfeng; Zheng Pu; Liu Rong; Yang Xiaofei; Yang Jian

2013-01-01

A fusion-fission hybrid conceptual reactor is established. It consists of a DT neutron source and a spherical shell of depleted uranium and hydrogen lithium. The tritium production rate (TPR) distribution in the conceptual reactor was measured by DT neutrons using two sets of lithium glass detectors with different thicknesses in the hole in the vertical direction with respect to the D + beam of the Cockcroft-Walton neutron generator in direct current mode. The measured TPR distribution is compared with the calculated results obtained by the three-dimensional Monte Carlo code MCNP5 and the ENDF/B-Ⅵ data file. The discrepancy between the measured and calculated values can be attributed to the neutron data library of the hydrogen lithium lack S(α, β) thermal scattering model, so we show that a special database of low-energy and thermal neutrons should be established in the physics design of fusion-fission hybrid reactors. (authors)

Mo-99 production on a LEU solution reactor

International Nuclear Information System (INIS)

Brown, R.W.; Thome, L.A.; Khvostionov, V.Y.

2005-01-01

A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO 2 SO 4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

Modeling issues associated with production reactor safety assessment

International Nuclear Information System (INIS)

Stack, D.W.; Thomas, W.R.

1990-01-01

This paper describes several Probabilistic Safety Assessment (PSA) modeling issues that are related to the unique design and operation of the production reactors. The identification of initiating events and determination of a set of success criteria for the production reactors is of concern because of their unique design. The modeling of accident recovery must take into account the unique operation of these reactors. Finally, a more thorough search and evaluation of common-cause events is required to account for combinations of unique design features and operation that might otherwise not be included in the PSA. It is expected that most of these modeling issues also would be encountered when modeling some of the other more unique reactor and nonreactor facilities that are part of the DOE nuclear materials production complex. 9 refs., 2 figs

Radioisotope tracer study in an aniline production reactor

International Nuclear Information System (INIS)

Pant, H.J.; Yelgoankar, V.N.; Mendhekar, G.N.

1995-01-01

A radioisotope tracer study was carried out in an aniline production reactor to investigate the cause of poor heat transfer from tube side to shell side in an aniline production (ANPO) reactor. The results of the study indicated that more than 50% of the shell volume was reduced due to deposition of the process material (i.e. fouling) on the shell walls and may be the cause of poor heat transfer in the reactor. (author). 2 refs., 2 figs

Radiolytic production of chemical fuels in fusion reactor systems

Energy Technology Data Exchange (ETDEWEB)

Fish, J D

1977-06-01

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

Radiolytic production of chemical fuels in fusion reactor systems

International Nuclear Information System (INIS)

Fish, J.D.

1977-06-01

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

Medical Radioisotopes Production Without A Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Van der Keur, H.

2010-05-15

This report is answering the key question: Is it possible to ban the use of research reactors for the production of medical radioisotopes? Chapter 2 offers a summarized overview on the history of nuclear medicine. Chapter 3 gives an overview of the basic principles and understandings of nuclear medicine. The production of radioisotopes and its use in radiopharmaceuticals as a tracer for imaging particular parts of the inside of the human body (diagnosis) or as an agent in radiotherapy. Chapter 4 lists the use of popular medical radioisotopes used in nuclear imaging techniques and radiotherapy. Chapter 5 analyses reactor-based radioisotopes that can be produced by particle accelerators on commercial scale, other alternatives and the advantages of the cyclotron. Chapter 6 gives an overview of recent developments and prospects in worldwide radioisotopes production. Chapter 7 presents discussion, conclusions and recommendations, and is answering the abovementioned key question of this report: Is it possible to ban the use of a nuclear reactor for the production of radiopharmaceuticals? Is a safe and secure production of radioisotopes possible?.

The effective management of medical isotope production in research reactors

International Nuclear Information System (INIS)

Drummond, D.T.

1993-01-01

During the 50-yr history of the use of radioisotopes for medical applications, research reactors have played a pivotal role in the production of many if not most of the key products. The marriage between research reactors and production operations is subject to significant challenges on two fronts. The medical applications of the radioisotope products impose some unique constraints and requirements on the production process. In addition, the mandates and priorities of a research reactor are not always congruent with the demands of a production environment. This paper briefly reviews the historical development of medical isotope production, identifies the unique challenges facing this endeavor, and discusses the management of the relationship between the isotope producer and the research reactor operator. Finally, the key elements of a successful relationship are identified

The Modular Helium Reactor for Hydrogen Production

International Nuclear Information System (INIS)

E. Harvego; M. Richards; A. Shenoy; K. Schultz; L. Brown; M. Fukuie

2006-01-01

For electricity and hydrogen production, an advanced reactor technology receiving considerable international interest is a modular, passively-safe version of the high-temperature, gas-cooled reactor (HTGR), known in the U.S. as the Modular Helium Reactor (MHR), which operates at a power level of 600 MW(t). For hydrogen production, the concept is referred to as the H2-MHR. Two concepts that make direct use of the MHR high-temperature process heat are being investigated in order to improve the efficiency and economics of hydrogen production. The first concept involves coupling the MHR to the Sulfur-Iodine (SI) thermochemical water splitting process and is referred to as the SI-Based H2-MHR. The second concept involves coupling the MHR to high-temperature electrolysis (HTE) and is referred to as the HTE-Based H2-MHR

JENDL special purpose file

International Nuclear Information System (INIS)

Nakagawa, Tsuneo

1995-01-01

In JENDL-3,2, the data on all the reactions having significant cross section over the neutron energy from 0.01 meV to 20 MeV are given for 340 nuclides. The object range of application extends widely, such as the neutron engineering, shield and others of fast reactors, thermal neutron reactors and nuclear fusion reactors. This is a general purpose data file. On the contrary to this, the file in which only the data required for a specific application field are collected is called special purpose file. The file for dosimetry is a typical special purpose file. The Nuclear Data Center, Japan Atomic Energy Research Institute, is making ten kinds of JENDL special purpose files. The files, of which the working groups of Sigma Committee are in charge, are listed. As to the format of the files, ENDF format is used similarly to JENDL-3,2. Dosimetry file, activation cross section file, (α, n) reaction data file, fusion file, actinoid file, high energy data file, photonuclear data file, PKA/KERMA file, gas production cross section file and decay data file are described on their contents, the course of development and their verification. Dosimetry file and gas production cross section file have been completed already. As for the others, the expected time of completion is shown. When these files are completed, they are opened to the public. (K.I.)

Nuclear reactor structural material forming less radioactive corrosion product

International Nuclear Information System (INIS)

Nakazawa, Hiroshi.

1988-01-01

Purpose: To provide nuclear reactor structural materials forming less radioactive corrosion products. Constitution: Ni-based alloys such as inconel alloy 718, 600 or inconel alloy 750 and 690 having excellent corrosion resistance and mechanical property even in coolants at high temperature and high pressure have generally been used as nuclear reactor structural materials. However, even such materials yield corrosion products being attacked by coolants circulating in the nuclear reactor, which produce by neutron irradiation radioactive corrosion products, that are deposited in primary circuit pipeways to constitute exposure sources. The present invention dissolves dissolves this problems by providing less activating nuclear reactor structural materials. That is, taking notice on the fact that Ni-58 contained generally by 68 % in Ni changes into Co-58 under irradiation of neutron thereby causing activation, the surface of nuclear reactor structural materials is applied with Ni plating by using Ni with a reduced content of Ni-58 isotopes. Accordingly, increase in the radiation level of the nuclear reactor structural materials can be inhibited. (K.M.)

Photofission observations in reactor environments using selected fission-product yields

International Nuclear Information System (INIS)

Gold, R.; Ruddy, F.H.; Roberts, J.H.

1982-01-01

A new method for the observation of photofission in reactor environments is advanced. It is based on the in-situ observation of fission product yield. In fact, at a given in-situ reactor location, the fission product yield is simply a weighted linear combination of the photofission product yield, Y/sub gamma/, and the neutron induced fission product yield, Y/sub n. The weight factors arising in this linear combination are the photofission fraction and neutron induced fission fraction, respectively. This method can be readily implemented with established techniques for measuring in-situ reactor fission product yield. For example, one can use the method based on simultaneous irradiation of radiometric (RM) and solid state track recorder (SSTR) fission monitors. The sensitivity and accuracy and current knowledge of fission product yields. Unique advantages of this method for reactor applications are emphasized

Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hho jung; Seul, K W; Ahn, S K; Bang, Y S; Park, D G; Kim, B K; Kim, W S; Lee, J H; Kim, W K; Shim, T M; Choi, H S; Ahn, H J; Jung, D W; Kim, G I; Park, Y M; Lee, Y J [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

1997-07-01

The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

The experimental nuclear reactor: AQUILON

International Nuclear Information System (INIS)

Girard, Y.; Koechlin, J.C.; Moreau, J.M.

1958-01-01

'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [fr

Radionuclide characterization of graphite stacks from plutonium production reactors of the Siberian group of chemical enterprises

International Nuclear Information System (INIS)

Bushuev, A.V.; Verzilov, Yu.M.; Zubarev, V.N.

2001-01-01

The residual radionuclide concentrations and distributions in graphite from moderator stack of plutonium production reactors at Tomsk-7 have been investigated. It was found that the dominant activity of graphite is 14 C. To gain information on surface and volume contamination of graphite blocks from the moderator stack, the special sets of samples were collected and assayed. The schemes are proposed for evaluation of individual radionuclide inventories together with results of the evaluations performed. (author)

Special issue on ageing management and long term operation of light water reactors

International Nuclear Information System (INIS)

Kashima, Koichi; Kanno, Masanori; Kimura, Itsurou

2008-01-01

Ageing management and long term operation of light water reactors became important in Japan and relevant research on physical and materials ageing has been carried out among organizations of government, academia and industry and establishment of technical standards and guideline based on the results is under way. The Japan Energy Policy Institute (JEPI) issued a special number discussing this theme, which consisted of ten reports of experts describing these activities. Main topics were ageing evaluation of reactor components due to neutron irradiation embrittlement and stress corrosion cracking, regulatory evaluation of deteriorations due to ageing with technical information basis, technology development in the area of inspection/monitoring, ageing evaluation and preventive maintenance/repairs, nuclear power plant life management of electric utilities, and advancement of reactor maintenance and inspection. (T. Tanaka)

Production capabilities in US nuclear reactors for medical radioisotopes

Energy Technology Data Exchange (ETDEWEB)

Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr. (Oak Ridge National Lab., TN (United States)); Schenter, R.E. (Westinghouse Hanford Co., Richland, WA (United States))

1992-11-01

The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted.

Production capabilities in US nuclear reactors for medical radioisotopes

International Nuclear Information System (INIS)

Mirzadeh, S.; Callahan, A.P.; Knapp, F.F. Jr.; Schenter, R.E.

1992-11-01

The availability of reactor-produced radioisotopes in the United States for use in medical research and nuclear medicine has traditionally depended on facilities which are an integral part of the US national laboratories and a few reactors at universities. One exception is the reactor in Sterling Forest, New York, originally operated as part of the Cintichem (Union Carbide) system, which is currently in the process of permanent shutdown. Since there are no industry-run reactors in the US, the national laboratories and universities thus play a critical role in providing reactor-produced radioisotopes for medical research and clinical use. The goal of this survey is to provide a comprehensive summary of these production capabilities. With the temporary shutdown of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) in November 1986, the radioisotopes required for DOE-supported radionuclide generators were made available at the Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR). In March 1988, however, the HFBR was temporarily shut down which forced investigators to look at other reactors for production of the radioisotopes. During this period the Missouri University Research Reactor (MURR) played an important role in providing these services. The HFIR resumed routine operation in July 1990 at 85 MW power, and the HFBR resumed operation in June 1991, at 30 MW power. At the time of the HFBR shutdown, there was no available comprehensive overview which could provide information on status of the reactors operating in the US and their capabilities for radioisotope production. The obvious need for a useful overview was thus the impetus for preparing this survey, which would provide an up-to-date summary of those reactors available in the US at both the DOE-funded national laboratories and at US universities where service irradiations are currently or expected to be conducted

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

Revisiting homogeneous suspension reactors for production of radioisotopes

International Nuclear Information System (INIS)

Pasqualini, E.E.

2010-01-01

Some 50 years ago in Geneva Conferences I, II and III (1955. 1958 and 1964) on the Peaceful Uses of Atomic Energy, and also in Vienna Symposium on Reactor Experiments (1961), several papers where presented by different countries referring to advances in homogeneous suspension reactors. In particular the Dutch KEMA Suspension Test Reactor (KSTR) was developed, built and successfully operated in the sixties and seventies. It was a 1MWth reactor in which a suspension (6 microns spheres) of mixed UO 2 /ThO 2 in light water was circulated in a closed loop through a sphere-shaped vessel. One of the basic ideas on these suspension reactors was to apply the fission recoil separation effect as a means of purification of the fuel: the non-volatile fission products can be adsorbed in dispersed active charcoal and removed from the liquid. Undoubtedly, this method can present some advantages and better yields for the production of Mo-99 and other short lived radioisotopes, since they have to be extracted from a liquid in which practically no uranium is present. Details are mentioned of the different aspects that have been taken into account and which ones could be added in the corresponding actualization of suspension reactors for radioisotope production. In recent years great advances have been made in nanotechnology that can be used in the tailoring of fuel particles and adsorbent media. Recently, in CNEA Buenos Aires, a new facility has been inaugurated and is being equipped and licensed for laboratory experiments and preparative synthesis of nuclear nanoparticles. RA-6 and RA-3 experimental reactors in Argentina can be used for in-pile testing. (author)

Integral measurement of fission products capture in fast breeder reactors

International Nuclear Information System (INIS)

Martin Deidier, Loick.

1979-12-01

For the SUPERPHENIX reactor project, it was necessary to know fission products capture with about 10% accuracy in the fast breeder reactor spectra. In this purpose, integral measurements have been carried out on the main separated products by different experimental technics (oscillation, activation and irradiation methods), but particularly on irradiated fuel pins from RAPSODIE and PHENIX reactors in order to directly obtain total effect of fission products. Same tendencies have been observed for both enriched uranium fuel and LMFBR characteristic plutonium fuel. All experimental results have been introduced in CARNAVAL cross section set [fr

Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

International Nuclear Information System (INIS)

Anon.

1981-01-01

The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

Iiyoshi, A.; Momota, H.; Motojima, O.

1994-01-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

International Nuclear Information System (INIS)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author)

Innovative energy production in fusion reactors

Science.gov (United States)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

Innovative energy production in fusion reactors

Energy Technology Data Exchange (ETDEWEB)

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

1993-10-01

Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are (a) traveling wave direct energy conversion of 14.7 MeV protons, (b) cusp type direct energy conversion of charged particles, (c) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas, and (d) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising. (author).

An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

International Nuclear Information System (INIS)

Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

2006-01-01

The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

Aspects of 238Pu production in the experimental fast reactor JOYO

International Nuclear Information System (INIS)

Osaka, Masahiko; Koyama, Shin-ichi; Tanaka, Kenya; Itoh, Masahiko; Saito, Masaki

2005-01-01

Experimental determination of 238 Pu in 237 Np samples irradiated in the experimental fast reactor JOYO was done as part of the demonstration of 238 Pu production from 237 Np in fast reactors within the framework of the protected Pu production project, which aims at reinforcement of proliferation resistance of Pu by increasing the 238 Pu isotopic ratio. 238 Pu production amount in the irradiated 237 Np samples was determined by a radioanalytical technique. Aspects of 238 Pu production were examined on the basis of the present radioanalysis. The 238 Pu production amount depends on the neutron spectrum which can range from that of a typical fast reactor to a nearly epi-thermal spectrum. It is concluded that the fast reactor has not only high potential for use in protected Pu production, but also as an incinerator for excess Pu

Use of tower reactors for continuous ethanol production

Directory of Open Access Journals (Sweden)

M.C. Viegas

2002-04-01

Full Text Available The purpose of this work was to develop a continuous fermentation system operating with a tower reactor using some flocculent yeast strains isolated from an industrial process. The strain was an used in the trial of the proposed system, composed of two serial glass tower reactor. The effects of the following variables were studied on the yield and productivity of the system: total reducing sugar (TRS, concentration in feeding, recycle flow in the second reactor, residence time and diameter/height ratio of the reactors. It was observed that the TRS concentration in feeding and residence time is the variables that interfere most with the productivity of the system. Yield was not affected by any of the variables within the range of values studied. All trials were performed according to a factorial experimental design (making up a total of 19 trials and the results were evaluated by response surface.

Synfuels production from fusion reactors

International Nuclear Information System (INIS)

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

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

Fuel development for reactors of new generation in Ukraine

International Nuclear Information System (INIS)

Odeychuk, N.P.

2006-01-01

Full text: On the background of critical situation in traditional power engineering due to deficiency of organic fuel, physical and moral ageing of the of thermal power stations equipment and their harmful influence on the ecology of environment, the nuclear engineering works stably enough and, by keeping all safety measures, is the most non-polluting energy source. In Ukraine the atomic engineering became one of main sources of energy production and is the important factor of guarantee the power engineering independence of the state. The main center on development of the components of nuclear reactors active zones is the National scientific center K harkov institute of Physics and Technology . The significant place in institutes' investigations was occupied with works on creation the constructional materials and nuclear fuel for heavy water reactors E-circumflexS-150, OR-1000, OR-2000, light water reactors WWER-1000 and RBMK-1500, high-temperature gas cooled reactors ABTU and HTGR, gas reactors on fast neutrons BGR and BRGD, and also the reactor - converter ROMASHKA and other special reactors of special assignment. Radiation tests and post-irradiation research confirm intended material-study, technological and design decisions and fuel elements capacity work on the whole. Nevertheless, by the present conditions, it is necessary to pay special attention to development of the new, safe guaranteed nuclear energy sources. In Ukraine proceed works on research and development of new safe nuclear reactors: basing the underground nuclear thermal power stations; development the reactors with managed chain reaction of nucleus division in an active zone with the help of an external source of neutrons; power thermonuclear installations; high-temperature helium reactors which are especially actual now from the point of view of the hydrogen production; the advanced pressure water reactors, heavy water reactors. In the paper also discussed the state of works in Ukraine on fuel

Production of energy in a high temperature reactor

International Nuclear Information System (INIS)

Anon.

1979-01-01

The cooling gas having left the reactor core is fed to a generator for direct production of current from the kinetic energy. Afterwards the gas is fed to a heat exchanger for cooling, then compressed and refed to the reactor core. The method further comprises that one part of the energy of the fission material is directly converted to electric energy in the reactor core, whereas the other part of the energy of the fission material is impressed upon the cooling gas. According to the invention the cooling gas when entering the reactor is first fed to that part of the reactor core which serves as a thermoionic or thermoelectric transducer. Afterwards the cooling gas is fed to the remaining part of the reactor gas. (P.K.)

Savannah River Site production reactor safety analysis report

International Nuclear Information System (INIS)

1996-01-01

The process water system (PWS) is designed to remove heat produced in the reactor from the fission process, gamma radiation absorption, and fission product decay. Heat removal is accomplished by circulating heavy water through the reactor. Cooling is provided for fuel assemblies, target assemblies, control rods, bulk moderator, deflector plate, reactor tank, and reactor structural components. Approximately 90% of the heat load is generated in the fuel and target assemblies, 5% in the moderator, and 5% in the shielding. In addition to serving as the-heat transfer medium, the process water moderates neutrons produced by fission in the fuel. D 2 O is used in this application because of its favorable moderating and neutron capture properties, which result in high neutron efficiency and reactor productivity. The PWS piping and components also provide a high-integrity leak barrier against loss of moderator and the radioactive fission and corrosion products. Components of the PWS are located in the reactor building between the -40-foot elevation and the 0-foot elevation. Specific locations include the process room, heat exchanger bay, motor rooms, and pump rooms. The system diagram is shown in Figure 5.1-2. PWS design data are presented in Table 5.1-1. The PWS consists of six parallel heat transfer loops. In each loop, approximately 25,000 gpm of D 2 O is circulated from one of six outlet nozzles in the bottom of the reactor tank through a motor-operated valve (MOV) to the suction side of the process water pump. Each pump is driven by an AC motor and a DC motor through a gear reducer unit. A 3-ton flywheel on the drive shaft of the AC motor provides gradual flow coastdown when power is lost. During reactor operation, the DC motors are operated continuously from the diesel generator sets as backup to the AC motors. Following shutdown, the DC motors are operated to provide adequate circulation and core cooling

Decreasing methane production in hydrogenogenic UASB reactors fed with cheese whey

International Nuclear Information System (INIS)

Carrillo-Reyes, Julián; Celis, Lourdes B.; Alatriste-Mondragón, Felipe; Razo-Flores, Elías

2014-01-01

One of the problems in fermentative hydrogen producing reactors, inoculated with pre-treated anaerobic granular sludge, is the eventual methane production by hydrogen-consuming methanogens. In this study, strategies such as reduction of pH and HRT, organic shock loads and repeated biomass heat treatment were applied to hydrogenogenic UASB reactors fed with cheese whey, that showed methane production after certain time of continuous operation (between 10 and 60 days). The reduction of pH to 4.5 not only decreased methane production but also hydrogen production. Organic shock load (from 20 to 30 g COD/L-d) was the more effective strategy to decrease the methane production rate (75%) and to increase the hydrogen production rate (172%), without stopping reactor operation. Repeated heat treatment of the granular sludge was the only strategy that inhibited completely methane production, leading to high volumetric hydrogen production rates (1.67 L H 2 /L-d), however this strategy required stopping reactor operation; in addition homoacetogenesis, another hydrogen-consuming pathway, was not completely inhibited. This work demonstrated that it was possible to control the methane activity in hydrogen producing reactors using operational strategies. - Highlights: • Operational strategies control methane in hydrogen production from cheese whey. • Organic shock load increased the hydrogen production rate. • Operation pH below 5 decreased both the hydrogen and methane production. • Second biomass heat treatment inhibits completely methanogenesis. • Homoacetogens play a negative role in fermentative hydrogen production

High temperature fast reactor for hydrogen production in Brazil

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Ono, Shizuca; Guimaraes, Lamartine N.F.

2008-01-01

The main nuclear reactors technology for the Generation IV, on development phase for utilization after 2030, is the fast reactor type with high temperature output to improve the efficiency of the thermo-electric conversion process and to enable applications of the generated heat in industrial process. Currently, water electrolysis and thermo chemical cycles using very high temperature are studied for large scale and long-term hydrogen production, in the future. With the possible oil scarcity and price rise, and the global warming, this application can play an important role in the changes of the world energy matrix. In this context, it is proposed a fast reactor with very high output temperature, ∼ 1000 deg C. This reactor will have a closed fuel cycle; it will be cooled by lead and loaded with nitride fuel. This reactor may be used for hydrogen, heat and electricity production in Brazil. It is discussed a development strategy of the necessary technologies and some important problems are commented. The proposed concept presents characteristics that meet the requirements of the Generation IV reactor class. (author)

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

Science.gov (United States)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.

Hydrogen production system coupled with high-temperature gas-cooled reactor (HTTR)

International Nuclear Information System (INIS)

Shiozawa, Shusaku

2003-01-01

On the HTTR program, R and D on nuclear reactor technology and R and D on thermal application technology such as hydrogen production and so on, are advanced. When carrying out power generation and thermal application such as hydrogen production and so on, it is, at first, necessary to supply nuclear heat safely, stably and in low cost, JAERI carries out some R and Ds on nuclear reactor technology using HTTR. In parallel to this, JAERI also carries out R and D for jointing nuclear reactor system with thermal application systems because of no experience in the world on high temperature heat of about 1,000 centigrade supplied by nuclear reactor except power generation, and R and D on thermochemical decomposition method IS process for producing hydrogen from water without exhaust of carbon dioxide. Here were described summaries on R and D on nuclear reactor technology, R and D on jointing technology using HTTR hydrogen production system, R and D on IS process hydrogen production, and comparison hydrogen production with other processes. (G.K.)

Biological hydrogen production by Clostridium acetobutylicum in an unsaturated flow reactor.

Science.gov (United States)

Zhang, Husen; Bruns, Mary Ann; Logan, Bruce E

2006-02-01

A mesophilic unsaturated flow (trickle bed) reactor was designed and tested for H2 production via fermentation of glucose. The reactor consisted of a column packed with glass beads and inoculated with a pure culture (Clostridium acetobutylicum ATCC 824). A defined medium containing glucose was fed at a flow rate of 1.6 mL/min (0.096 L/h) into the capped reactor, producing a hydraulic retention time of 2.1 min. Gas-phase H2 concentrations were constant, averaging 74 +/- 3% for all conditions tested. H2 production rates increased from 89 to 220 mL/hL of reactor when influent glucose concentrations were varied from 1.0 to 10.5 g/L. Specific H2 production rate ranged from 680 to 1270 mL/g glucose per liter of reactor (total volume). The H2 yield was 15-27%, based on a theoretical limit by fermentation of 4 moles of H2 from 1 mole of glucose. The major fermentation by-products in the liquid effluent were acetate and butyrate. The reactor rapidly (within 60-72 h) became clogged with biomass, requiring manual cleaning of the system. In order to make long-term operation of the reactor feasible, biofilm accumulation in the reactor will need to be controlled through some process such as backwashing. These tests using an unsaturated flow reactor demonstrate the feasibility of the process to produce high H2 gas concentrations in a trickle-bed type of reactor. A likely application of this reactor technology could be H2 gas recovery from pre-treatment of high carbohydrate-containing wastewaters.

Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Werner, James Elmer [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; McKellar, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Kennedy, John Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Biersdorf, John Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division

2017-04-01

The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

Measuring the productivity of university research reactors

International Nuclear Information System (INIS)

Voth, M.H.

1989-01-01

University Research Reactors (URRs) on 33 campuses in the United States provide valuable contributions to academic instruction and research programs. In most cases, there are no alternative diagnostic techniques to supplant the need for a reactor and associated facilities. Since URRs constitute a major financial commitment, it is important that they be operated in a productive manner. Productivity may be defined as the sum of new knowledge generated, existing knowledge transferred to others, and analytical services provided to assist in the generation of new knowledge; another definition of productivity is this sum expressed as a function of the cost incurred. In either case, a consistent measurement is difficult and more qualitative than quantitative. A uniform reporting system has been proposed that defines simplified categories through which meaningful comparisons can be performed

Large-scale hydrogen production using nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Ryland, D.; Stolberg, L.; Kettner, A.; Gnanapragasam, N.; Suppiah, S. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

For many years, Atomic Energy of Canada Limited (AECL) has been studying the feasibility of using nuclear reactors, such as the Supercritical Water-cooled Reactor, as an energy source for large scale hydrogen production processes such as High Temperature Steam Electrolysis and the Copper-Chlorine thermochemical cycle. Recent progress includes the augmentation of AECL's experimental capabilities by the construction of experimental systems to test high temperature steam electrolysis button cells at ambient pressure and temperatures up to 850{sup o}C and CuCl/HCl electrolysis cells at pressures up to 7 bar and temperatures up to 100{sup o}C. In parallel, detailed models of solid oxide electrolysis cells and the CuCl/HCl electrolysis cell are being refined and validated using experimental data. Process models are also under development to assess options for economic integration of these hydrogen production processes with nuclear reactors. Options for large-scale energy storage, including hydrogen storage, are also under study. (author)

The modular high-temperature gas-cooled reactor - a new production reactor

International Nuclear Information System (INIS)

Nulton, J.D.

1990-01-01

One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

Fourth annual progress report on special-purpose materials for magnetically confined fusion reactors

International Nuclear Information System (INIS)

1982-08-01

The scope of Special Purpose Materials covers fusion reactor materials problems other than the first-wall and blanket structural materials, which are under the purview of the ADIP, DAFS, and PMI task groups. Components that are considered as special purpose materials include breeding materials, coolants, neutron multipliers, barriers for tritium control, materials for compression and OH coils and waveguides, graphite and SiC, heat-sink materials, ceramics, and materials for high-field (>10-T) superconducting magnets. The Task Group on Special Purpose Materials has limited its concern to crucial and generic materials problems that must be resolved if magnetic-fusion devices are to succeed. Important areas specifically excluded include low-field (8-T) superconductors, fuels for hybrids, and materials for inertial-confinement devices. These areas may be added in the future when funding permits

Summary - Advanced high-temperature reactor for hydrogen and electricity production

International Nuclear Information System (INIS)

Forsberg, Charles W.

2001-01-01

Historically, the production of electricity has been assumed to be the primary application of nuclear energy. That may change. The production of hydrogen (H 2 ) may become a significant application. The technology to produce H 2 using nuclear energy imposes different requirements on the reactor, which, in turn, may require development of new types of reactors. Advanced High Temperature reactors can meet the high temperature requirements to achieve this goal. This alternative application of nuclear energy may necessitate changes in the regulatory structure

Bio-hydrogen production from molasses by anaerobic fermentation in continuous stirred tank reactor

Science.gov (United States)

Han, Wei; Li, Yong-feng; Chen, Hong; Deng, Jie-xuan; Yang, Chuan-ping

2010-11-01

A study of bio-hydrogen production was performed in a continuous flow anaerobic fermentation reactor (with an available volume of 5.4 L). The continuous stirred tank reactor (CSTR) for bio-hydrogen production was operated under the organic loading rates (OLR) of 8-32 kg COD/m3 reactor/d (COD: chemical oxygen demand) with molasses as the substrate. The maximum hydrogen production yield of 8.19 L/d was obtained in the reactor with the OLR increased from 8 kg COD/m3 reactor/d to 24 kg COD/m3 d. However, the hydrogen production and volatile fatty acids (VFAs) drastically decreased at an OLR of 32 kg COD/m3 reactor/d. Ethanoi, acetic, butyric and propionic were the main liquid fermentation products with the percentages of 31%, 24%, 20% and 18%, which formed the mixed-type fermentation.

Radiation hazards due to activated corrosion and neutron sputtering products in fusion reactor coolant and tritium breeding fluids

International Nuclear Information System (INIS)

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

1985-01-01

The accumulation of radioactive corrosion and neutron sputtering products on the surfaces of components in fusion reactor coolant and tritium breeding systems can cause significant personnel access problems. Remote maintenance techniques or special treatment may be required to limit the amount of radiation exposure to plant operational and maintenance personnel. A computer code, RAPTOR, has been developed to estimate the transport of this activated material throughout a fusion heat transfer and/or tritium breeding material loop. A method is devised which treats the components of the loop individually and determines the source rates, deposition and erosion rates, decay rates, and purification rates of these radioactive materials. RAPTOR has been applied to the MARS and Starfire conceptual reactor designs to determine the degree of the possible radiation hazard due to these products. Due to the very high corrosion release rate by HT-9 when exposed to LiPb in the MARS reactor design, the radiation fields surrounding the primary system will preclude direct contact maintenance even after shutdown. Even the removal of the radioactive LiPb from the system will not decrease the radiation fields to reasonable levels. The Starfire primary system will exhibit radiation fields similar to those found in present pressurized water reactors. (orig.)

Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

Energy Technology Data Exchange (ETDEWEB)

Rubindamayugi, M S.T.; Salakana, L K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

1998-12-31

The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

Biogas production from UASB and polyurethane carrier reactors treating sisal processing wastewater

Energy Technology Data Exchange (ETDEWEB)

Rubindamayugi, M.S.T.; Salakana, L.K.P. [Univ. of Dar es Salaam, Faculty of Science, Applied Microbiology Unit (Tanzania, United Republic of)

1997-12-31

The fundamental benefits which makes anaerobic digestion technology (ADT) attractive to the poor developing include the low cost and energy production potential of the technology. In this study the potential of using UASB reactor and Polyurethane Carrier Reactor (PCR) as pollution control and energy recovery systems from sisal wastewater were investigated in lab-scale reactors. The PCR demonstrated the shortest startup period, whereas the UASB reactor showed the highest COD removal efficiency 79%, biogas production rate (4.5 l biogas/l/day) and process stability than the PCR under similar HRT of 15 hours and OLR of 8.2 g COD/l/day. Both reactor systems became overloaded at HRT of 6 hours and OLR of 15.7 g COD/l/day, biogas production ceased and reactors acidified to pH levels which are inhibiting to methanogenesis. Based on the combined results on reactor performances, the UASB reactor is recommended as the best reactor for high biogas production and treatment efficiency. It was estimated that a large-scale UASB reactor can be designed under the same loading conditions to produce 2.8 m{sup 3} biogas form 1 m{sup 3} of wastewater of 5.16 kg COD/m{sup 3}. Wastewater from one decortication shift can produce 9,446 m{sup 3} og biogas. The energy equivalent of such fuel energy is indicated. (au)

Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

International Nuclear Information System (INIS)

Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

2008-01-01

In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

Safety of VVER-440 reactors

CERN Document Server

Slugen, Vladimir

2011-01-01

Safety of VVER-440 Reactors endeavours to promote an increase in the safety of VVER-440 nuclear reactors via the improvement of fission products limitation systems and the implementation of special non-destructive spectroscopic methods for materials testing. All theoretical and experimental studies performed the by author over the last 25 years have been undertaken with the aim of improving VVER-440 defence in depth, which is one of the most important principle for ensuring safety in nuclear power plants. Safety of VVER-440 Reactors is focused on the barrier system through which the safety pri

Fission product chemistry in severe nuclear reactor accidents

International Nuclear Information System (INIS)

Nichols, A.L.

1990-09-01

A specialist's meeting was held at JRC-Ispra from 15 to 17 January 1990 to review the current understanding of fission-product chemistry during severe accidents in light water reactors. Discussions focussed on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. Recommendations for future chemistry work were made covering the following areas: (a) fuel degradation and fission-product release, (b) transport and attenuation processes in the reactor coolant system, (c) containment chemistry (iodine behaviour and core-concrete interactions)

Reactor and method for production of nanostructures

Science.gov (United States)

Sunkara, Mahendra Kumar; Kim, Jeong H.; Kumar, Vivekanand

2017-04-25

A reactor and method for production of nanostructures, including metal oxide nanowires or nanoparticles, are provided. The reactor includes a regulated metal powder delivery system in communication with a dielectric tube; a plasma-forming gas inlet, whereby a plasma-forming gas is delivered substantially longitudinally into the dielectric tube; a sheath gas inlet, whereby a sheath gas is delivered into the dielectric tube; and a microwave energy generator coupled to the dielectric tube, whereby microwave energy is delivered into a plasma-forming gas. The method for producing nanostructures includes providing a reactor to form nanostructures and collecting the formed nanostructures, optionally from a filter located downstream of the dielectric tube.

Evaluation of the qualification of SPERT [Special Power Excursion Reactor Test] fuel for use in non-power reactors

International Nuclear Information System (INIS)

1987-08-01

This report summarizes the US Nuclear Regulatory Commission staff's evaluation of the qualification of the stainless-steel-clad uranium/oxide (UO 2 ) fuel pins for use in non-power reactors. The fuel pins were originally procured in the 1960's as part of the Special Power Excursion Reactor Test (SPERT) program. Argonne National Laboratory (ANL) examined 600 SPERT fuel pins to verify that the pins were produced according to specification and to assess their present condition. The pins were visually inspected under 6X magnification and by X-radiographic, destructive, and metallographic examinations. Spectrographic and chemical analyses were performed on the UO 2 fuel. The results of the qualification examinations indicated that the SPERT fuel pins meet the requirements of Phillips Specification No. F-1-SPT and have suffered no physical damage since fabrication. Therefore, the qualification results give reasonable assurance that the SPERT fuel rods are suitable for use in non-power reactors provided that the effects of thin-wall defects in the region of the upper end cap and low-density fuel pellets are evaluated for the intended operating conditions. 1 ref., 4 figs., 11 tabs

Pulsed fusion reactors

International Nuclear Information System (INIS)

1975-01-01

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

FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

International Nuclear Information System (INIS)

Auterinen, I.; Salmenhaara, S.E.J. . Author

2004-01-01

The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

Some methods of failed fuel element detection in water cooled reactors

International Nuclear Information System (INIS)

Strindehag, O.M.

1976-01-01

The methods are surveyed using fission products released in the coolant for the detection of failed fuel elements in water cooled reactors. The classification of the detection methods is made with respect to fission product detection in the coolant and to gaseous fission product detection. The detection systems are listed used for the AGESTA power reactor and for the experimental loops of the RA research reactor based on the detection of either gaseous fission products or gaseous daughter products. The AGESTA reactor detection systems using electrostatic precipitators consist of five precipitator channels of which three are intended for detection and two for localization. A special detection unit was developed for the failed fuel element detection in the R-2 reactor experimental steam loop. Its description is listed. In the reactor pressurized-water loop a Cherenkov counter was used in the detection of fission products. An ion exchange monitor whose application is described was used in the total measurement of the main coolant flow in the AGESTA reactor. (J.P.)

Optimization of neutron flux distribution in Isotope Production Reactor

International Nuclear Information System (INIS)

Valladares, G.L.

1988-01-01

In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely the Vmod / Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO 2 ) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author) [pt

Maximum production rate optimization for sulphuric acid decomposition process in tubular plug-flow reactor

International Nuclear Information System (INIS)

Wang, Chao; Chen, Lingen; Xia, Shaojun; Sun, Fengrui

2016-01-01

A sulphuric acid decomposition process in a tubular plug-flow reactor with fixed inlet flow rate and completely controllable exterior wall temperature profile and reactants pressure profile is studied in this paper by using finite-time thermodynamics. The maximum production rate of the aimed product SO 2 and the optimal exterior wall temperature profile and reactants pressure profile are obtained by using nonlinear programming method. Then the optimal reactor with the maximum production rate is compared with the reference reactor with linear exterior wall temperature profile and the optimal reactor with minimum entropy generation rate. The result shows that the production rate of SO 2 of optimal reactor with the maximum production rate has an increase of more than 7%. The optimization of temperature profile has little influence on the production rate while the optimization of reactants pressure profile can significantly increase the production rate. The results obtained may provide some guidelines for the design of real tubular reactors. - Highlights: • Sulphuric acid decomposition process in tubular plug-flow reactor is studied. • Fixed inlet flow rate and controllable temperature and pressure profiles are set. • Maximum production rate of aimed product SO 2 is obtained. • Corresponding optimal temperature and pressure profiles are derived. • Production rate of SO 2 of optimal reactor increases by 7%.

Development of small and medium reactors for power and heat production

International Nuclear Information System (INIS)

Becka, J.

1978-01-01

Data are given on the current state of development of small and medium-power reactors designed mainly for electric power production in small power grids, for heat production for small- and medium-power desalination plants with possible electric power generation, for process steam production and heat development for district heating systems, again combined with electric power generation, and for propelling big and fast passenger ships. A diagram is shown of the primary system of an integrated PWR derived from the Otto Hahn reactor. The family is listed of the standard sizes of the integral INTERATOM company pressurized water reactors. Also listed are the specifications and design of CAS 2CG and AS 3G type reactors used mainly for long-distance heating systems. (J.B.)

Utilization of fast reactor excess neutrons for burning long-lived fission products

International Nuclear Information System (INIS)

Kawashima, K.; Kobayashi, K.; Kaneto, K.

1995-01-01

An evaluation is made on a large MOX fuel fast reactor's capability of burning long lived fission product Tc-99, which dominates the long term radiotoxicity of the high level radioactive waste. The excess neutrons generated in the fast reactor core are utilized to transmute Tc-99 to stable isotopes due to neutron capture reaction. The fission product target assemblies which consist of Tc-99 are charged to the reactor core periphery. The fission product target neutrons are moderated to a great deal to pursue the possibility of enhancing the transmutation rate. Any impacts of loading the fission product target assemblies on the core nuclear performances are assessed. A long term Tc-99 accumulation scenario is considered in the mix of fission product burner fast reactor and non-burner LWRs. (author)

Experience of on-site disposal of production uranium-graphite nuclear reactor.

Science.gov (United States)

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.

NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

Energy Technology Data Exchange (ETDEWEB)

Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

2004-12-01

Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

Health physics aspects of activation products from fusion reactors

International Nuclear Information System (INIS)

Shoup, R.L.; Poston, J.W.; Easterly, C.E.; Jacobs, D.G.

1975-01-01

A review of the activation products from fusion reactors and their attendant impacts is discussed. This includes a discussion on their production, expected inventories, and the status of metabolic data on these products

Supplying the nuclear arsenal: Production reactor technology, management, and policy, 1942--1992

Energy Technology Data Exchange (ETDEWEB)

Carlisle, R.P.; Zenzen, J.M.

1994-01-01

This book focuses on the lineage of America`s production reactors, those three at Hanford and their descendants, the reactors behind America`s nuclear weapons. The work will take only occasional sideways glances at the collateral lines of descent, the reactor cousins designed for experimental purposes, ship propulsion, and electric power generation. Over the decades from 1942 through 1992, fourteen American production reactors made enough plutonium to fuel a formidable arsenal of more than twenty thousand weapons. In the last years of that period, planners, nuclear engineers, and managers struggled over designs for the next generation of production reactors. The story of fourteen individual machines and of the planning effort to replace them might appear relatively narrow. Yet these machines lay at the heart of the nation`s nuclear weapons complex. The story of these machines is the story of arming the winning weapon, supplying the nuclear arms race. This book is intended to capture the history of the first fourteen production reactors, and associated design work, in the face of the end of the Cold War.

Enviromental Development Plan: special nuclear materials production

International Nuclear Information System (INIS)

1980-07-01

This Environmental Development Plan includes the process steps and facilities necessary for the production of plutonium and tritium for Government needs and the production of some other radioactive materials that will be used for heat and radiation sources by domestic and international customers. The production reactors and the spent fuel processing plants and their effluents are discussed, but the defense wastes from them are treated in a separate EDP. The scope does not include transportation, decontamination and decommissioning, safeguards and security, or use of the SNM products

Specificities of micro-reactors for hydrogen production and purification

Energy Technology Data Exchange (ETDEWEB)

Mirodatos, C.; Dupont, N.; Germani, G.; Veen, A. C. ven; Schuurman, Y.

2005-07-01

Sustainable chemistry and exploitation of energy sources for the next decades requires considerable progress in process intensification. A development of new tools and equipments meeting the objectives of high efficiency, improved safety, compactness and low implementation costs is therefore subject of intensive research effort. Among the various scenarios tested in R and D, micro-structured reactors appear as a highly promising technology 1 and perspectives of mass production are already announced by technology providers 2. These reactors are based on assembly/stacking of micro structured plates or fibres. Due to their high heat and/or mass transfer, low pressure drop and good phase contacting, they sound particularly adapted to the large domain of hydrogen production by fuel reforming and purification. This presentation aims at outlining the state of the art, the advantages and drawbacks of using micro-structured reactors to intensify hydrogen production and purification. Two case studies will illustrate this approach: i) comparison between fixed bed and micro-structured reactor for the reforming of methanol into hydrogen and carbon oxides and ii) use of those devices in kinetic studies on the WGS reaction. (Author)

Synthesis of nanocrystalline Y2O3 in a specially designed atmospheric pressure radio frequency thermal plasma reactor

International Nuclear Information System (INIS)

Dhamale, G. D.; Mathe, V. L.; Bhoraskar, S. V.; Sahasrabudhe, S. N.; Ghorui, S.

2015-01-01

Synthesis of yttrium oxide nanoparticles in a specially designed radio frequency thermal plasma reactor is reported. Good crystallinity, narrow size distribution, low defect state concentration, high purity, good production rate, single-step synthesis, and simultaneous formation of nanocrystalline monoclinic and cubic phases are some of the interesting features observed. Synthesized particles are characterized through X-ray diffraction, transmission electron microscopy, scanning electron microscopy, Fourier transform infrared spectroscopy, thermo-luminescence (TL), and Brunauer–Emmett–Teller surface area analysis. Polymorphism of the nanocrystalline yttria is addressed in detail. Synthesis mechanism is explored through in-situ emission spectroscopy. Post-synthesis environmental effects and possible methods to eliminate the undesired phases are probed. Defect states are investigated through the study of TL spectra

Transmutation of nuclear waste in nuclear reactors

International Nuclear Information System (INIS)

Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

1996-03-01

The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

CFD optimization of continuous stirred-tank (CSTR) reactor for biohydrogen production.

Science.gov (United States)

Ding, Jie; Wang, Xu; Zhou, Xue-Fei; Ren, Nan-Qi; Guo, Wan-Qian

2010-09-01

There has been little work on the optimal configuration of biohydrogen production reactors. This paper describes three-dimensional computational fluid dynamics (CFD) simulations of gas-liquid flow in a laboratory-scale continuous stirred-tank reactor used for biohydrogen production. To evaluate the role of hydrodynamics in reactor design and optimize the reactor configuration, an optimized impeller design has been constructed and validated with CFD simulations of the normal and optimized impeller over a range of speeds and the numerical results were also validated by examination of residence time distribution. By integrating the CFD simulation with an ethanol-type fermentation process experiment, it was shown that impellers with different type and speed generated different flow patterns, and hence offered different efficiencies for biohydrogen production. The hydrodynamic behavior of the optimized impeller at speeds between 50 and 70 rev/min is most suited for economical biohydrogen production. Copyright 2010 Elsevier Ltd. All rights reserved.

Decommissioning planning and the assessment of alternatives for the Hanford production reactors

International Nuclear Information System (INIS)

Miller, C.E. Jr.; Potter, R.F.

1985-01-01

Several years ago, the US Department of Energy began assessing alternatives and planning the decommissioning of eight shut-down plutonium production reactors located on the DOE Hanford Site in Washington State. The first of these graphite-moderated, water-cooled, reactors was built and started up in 1944 as part of the World War II Manhattan Project. The last of them started up in 1955. The eight reactors each operated for 12 to 24 years, with all eight operating simultaneously for about 10 years. In the 1960's, production needs declined and the reactors were one-by-one permanently shut down, the last of them in 1971. (A ninth Hanford production reactor, N Reactor, was started up in 1963; it is still operating and is not within the scope of the decommissioning planning and alternatives assessment work reported in this paper). This paper provides an overview description of the decommissioning plan for the eight shut-down Hanford production reactors and their associated fuel storage basins. Included are descriptions of the decommissioning alternatives considered for the facilities, along with discussions of National Environmental Policy Act (NEPA) process activities applicable to the Hanford decommissioning work. The criteria used in assessing decommissioning alternatives and the assumptions used in the decommissioning planning are identified. 4 refs., 8 figs., 3 tabs

Helium production in mixed spectrum reactor-irradiated pure elements

International Nuclear Information System (INIS)

Kneff, D.W.; Oliver, B.M.; Skowronski, R.P.

1986-01-01

The objectives of this work are to apply helium accumulation neutron dosimetry to the measurement of neutron fluences and energy spectra in mixed-spectrum fission reactors utilized for fusion materials testing, and to measure helium generation rates of materials in these irradiation environments. Helium generation measurements have been made for several Fe, Cu Ti, Nb, Cr, and Pt samples irradiated in the mixed-spectrum High Flux Isotope Reactor (HFIR) and Oak Ridge Research Reactor (ORR) at the Oak Ridge National Laboratory. The results have been used to integrally test the ENDF/B-V Gas Production File, by comparing the measurements with helium generation predictions made by Argonne National Laboratory using ENDF/B-V cross sections and adjusted reactor spectra. The comparisons indicate consistency between the helium measurements and ENDF/B-V for iron, but cross section discrepancies exist for helium production by fast neutrons in Cu, Ti, Nb, and Cr (the latter for ORR). The Fe, Cu, and Ti work updates and extends previous measurements

Evaluation of Packed-Bed Reactor and Continuous Stirred Tank Reactor for the Production of Colchicine Derivatives

OpenAIRE

Dubey, Kashyap Kumar; Kumar, Dhirendra; Kumar, Punit; Haque, Shafiul; Jawed, Arshad

2013-01-01

Bioconversion of colchicine into its pharmacologically active derivative 3-demethylated colchicine (3-DMC) mediated by P450BM3 enzyme is an economic and promising strategy for the production of this inexpensive and potent anticancer drug. Continuous stirred tank reactor (CSTR) and packed-bed reactor (PBR) of 3 L and 2 L total volumes were compared for the production of 3-demethylated colchicine (3-DMC) a colchicine derivative using Bacillus megaterium MTCC*420 under aerobic conditions. Statis...

Production of Radioisotopes in Pakistan Research Reactor: Past, Present and Future

International Nuclear Information System (INIS)

Mushtaq, A.

2013-01-01

Radioisotope production to service different sectors of economic significance constitutes an important ongoing activity of many national nuclear programs. Radioisotopes, formed by nuclear reactions on targets in a reactor or cyclotron, require further processing in almost all cases to obtain them in a form suitable for use. The availability of short-lived radionuclides from radionuclide generators provides an inexpensive and convenient alternative to in-house radioisotope production facilities such as cyclotrons and reactors. The reactor offers large volume for irradiation, simultaneous irradiation of several samples, economy of production and possibility to produce a wide variety of radioisotopes. The accelerator-produced isotopes relatively constitute a smaller percentage of total use. (author)

Advanced nuclear reactor types and technologies

International Nuclear Information System (INIS)

Ignatiev, V.; Devell, L.

1995-01-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

Advanced nuclear reactor types and technologies

Energy Technology Data Exchange (ETDEWEB)

Ignatiev, V [ed.; Feinberg, O; Morozov, A [Russian Research Centre ` Kurchatov Institute` , Moscow (Russian Federation); Devell, L [Studsvik Eco and Safety AB, Nykoeping (Sweden)

1995-07-01

The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

Fission product behavior in the Molten Salt Reactor Experiment

International Nuclear Information System (INIS)

Compere, E.L.; Kirslis, S.S.; Bohlmann, E.G.; Blankenship, F.F.; Grimes, W.R.

1975-10-01

Essentially all the fission product data for numerous and varied samples taken during operation of the Molten Salt Reactor Experiment or as part of the examination of specimens removed after particular phases of operation are reported, together with the appropriate inventory or other basis of comparison, and relevant reactor parameters and conditions. Fission product behavior fell into distinct chemical groups. Evidence for fission product behavior during operation over a period of 26 months with 235 U fuel (more than 9000 effective full-power hours) was consistent with behavior during operation using 233 U fuel over a period of about 15 months (more than 5100 effective full-power hours)

Comparative Analysis of Hydrogen Production Methods with Nuclear Reactors

International Nuclear Information System (INIS)

Morozov, Andrey

2008-01-01

Hydrogen is highly effective and ecologically clean fuel. It can be produced by a variety of methods. Presently the most common are through electrolysis of water and through the steam reforming of natural gas. It is evident that the leading method for the future production of hydrogen is nuclear energy. Several types of reactors are being considered for hydrogen production, and several methods exist to produce hydrogen, including thermochemical cycles and high-temperature electrolysis. In the article the comparative analysis of various hydrogen production methods is submitted. It is considered the possibility of hydrogen production with the nuclear reactors and is proposed implementation of research program in this field at the IPPE sodium-potassium eutectic cooling high temperature experimental facility (VTS rig). (authors)

Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

International Nuclear Information System (INIS)

Anghaie, S.; Smith, B.

2004-01-01

This paper presents results of a conceptual design study involving gas and vapour core reactors (G/VCR) with a combined scheme to generate hydrogen and power. The hydrogen production schemes include high temperature electrolysis as well as two dominant thermochemical hydrogen production processes. Thermochemical hydrogen production processes considered in this study included the calcium-bromine process and the sulphur-iodine processes. G/VCR systems are externally reflected and moderated nuclear energy systems fuelled by stable uranium compounds in gaseous or vapour phase that are usually operated at temperatures above 1500 K. A gas core reactor with a condensable fuel such as uranium tetrafluoride (UF 4 ) or a mixture of UF 4 and other metallic fluorides (BeF 2 , LiF, KF, etc.) is commonly known as a vapour core reactor (VCR). The single most relevant and unique feature of gas/vapour core reactors is that the functions of fuel and coolant are combined into one. The reactor outlet temperature is not constrained by solid fuel-cladding temperature limits. The maximum fuel/working fluid temperature in G/VCR is only constrained by the reactor vessel material limits, which is far less restrictive than the fuel clad. Therefore, G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Gas and vapour fuel reactors feature very low fuel inventory and fully integrated fuel cycle that provide for exceptional sustainability and safety characteristics. With respect to fuel utilisation, there is no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to the flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from completely burner to breeder is achievable. The continuous recycling of fuel in G/VCR systems allow for complete burning of actinides without removing and reprocessing of the fuel. The only waste products at the back

Plutonium-239 production rate study using a typical fusion reactor

International Nuclear Information System (INIS)

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

2008-01-01

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

A novel spiral reactor for biodiesel production in supercritical ethanol

International Nuclear Information System (INIS)

Farobie, Obie; Sasanami, Kazuma; Matsumura, Yukihiko

2015-01-01

Highlights: • A novel spiral reactor for biodiesel production in supercritical ethanol was proposed. • The spiral reactor employed in this study successfully recovered heat. • The effects of temperature and time on FAEE yield were investigated. • FAEE yield as high as 0.937 mol/mol was obtained at 350 °C after 30 min. • The second-order kinetic model expressed the experimental yield well. - Abstract: A spiral reactor is proposed as a novel reactor design for biodiesel production under supercritical conditions. Since the spiral reactor serves as a heat exchanger, it offers the advantage of reduced apparatus space compared to conventional supercritical equipment. Experimental investigations were carried out at reaction temperatures of 270–400 °C, pressure of 20 MPa, oil-to-ethanol molar ratio of 1:40, and reaction times of 3–30 min. An FAEE yield of 0.937 mol/mol was obtained in a short reaction time of 30 min at 350 °C and oil-to-ethanol molar ratio of 1:40 under a reactor pressure of 20 MPa. The spiral reactor was not only as effective as conventional reactor in terms of transesterification reactor but also was superior in terms of heat recovery. A second-order kinetic model describing the transesterification of canola oil in supercritical ethanol was proposed, and the reaction was observed to follow Arrhenius behavior. The corresponding reaction rate constants and the activation energies as well as pre-exponential factors were determined

Testing plutonium fuel assembly production for fast-neutron reactors

International Nuclear Information System (INIS)

Nougues, B.; Benhamou, A.; Bertothy, G.; Lepetit, H.

1975-01-01

The main characteristics of plutonium fuel elements for fast breeder reactors justify specific test procedures and special techniques. The specific tests relating to the Pu content consist of Pu enrichment and distribution tests, determination of the O/M ratio and external contamination tests. The specific tests performed on fuel configuration are: testing of sintered pellet diameter, testing of pin welding and checking of internal assmbly [fr

Markets for reactor-produced non-fission radioisotopes

International Nuclear Information System (INIS)

Bennett, R.G.

1995-01-01

Current market segments for reactor produced radioisotopes are developed and reported from a review of current literature. Specific radioisotopes studied in is report are the primarily selected from those with major medical or industrial markets, or those expected to have strongly emerging markets. Relative market sizes are indicated. Special emphasis is given to those radioisotopes that are best matched to production in high flux reactors such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. A general bibliography of medical and industrial radioisotope applications, trends, and historical notes is included

Preparation of a primary target for the production of fission products in a nuclear reactor

International Nuclear Information System (INIS)

Arino, H.; Cosolito, F.J.; George, K.D.; Thornton, A.K.

1976-01-01

A primary target for the production of fission products in a nuclear reactor, such as uranium or plutonium fission products, is comprised of an enclosed, cylindrical vessel, preferably comprised of stainless steel, having a thin, continuous, uniform layer of fissionable material, integrally bonded to its inner walls and a port permitting access to the interior of the vessel. A process is also provided for depositing uranium material on to the inner walls of the vessel. Upon irradiation of the target with neutrons from a nuclear reactor, radioactive fission products, such as molybdenum-99, are formed, and thereafter separated from the target by the introduction of an acidic solution through the port to dissolve the irradiated inner layer. The irradiation and dissolution are thus effected in the same vessel without the necessity of transferring the fissionable material and fission products to a separate chemical reactor. Subsequently, the desired isotopes are extracted and purified. Molybdenum-99 decays to technetium-99m which is a valuable medical diagnostic radioisotope. 3 claims, 1 drawing figure

Study on hydrogen production using the fast breeder reactors (FBR)

International Nuclear Information System (INIS)

Kani, Yoshio

2003-01-01

As the fast breeder reactor (FBR) can effectively convert uranium-238 difficult to carry out nuclear fission at thermal neutron reactors to nuclear fissionable plutonium-239 to use it remarkable upgrading of application on uranium can be performed, to be expected for sustainable energy source. And, by reuse minor actinides of long half-life nuclides in reprocessed high level wasted solutions for fuels of nuclear reactors, reduction of radioactive poison based on high level radioactive wastes was enabled. As high temperature of about 800 centigrade was required on conventional hydrogen production, by new hydrogen production technique even at operation temperature of sodium-cooled FBR it can be enabled. Here were described for new hydrogen production methods applicable to FBR on palladium membrane hydrogen separation method carrying out natural gas/steam modification at reaction temperature of about 500 centigrade, low temperature thermo-chemical method expectable simultaneous simplification of production process, and electrolysis method expected on power load balancing. (G.K.)

Production reactor productivity improvement plan

International Nuclear Information System (INIS)

Leitz, E.E.

1980-01-01

The N Reactor complex, which is operated by UNC for DOE, is a unique facility and as such it is difficult to transfer technological developments and management innovations directly to the N Reactor operations. Therefore the approach to implementing an effective program was to start with the general systems philosophy and then progress into using those specific analytical and management techniques applicable to the unique situation (technologically and administratively) which existed at the N Reactor plant

Technology development and production of elongated shell for reactor vessel active zone of WWER-TOI project from steel 15Cr2NiMoVN class 1

International Nuclear Information System (INIS)

Shklyaev, S.Eh.; Titova, T.I.; Ratushev, D.V.; Shul'gan, N.A.; Eroshkin, S.B.; Durynin, V.A.; Efimov, S.V.; Dub, V.S.; Kulikov, A.P.; Romashkin, A.N.

2015-01-01

Production process for the elongated shell blank of the active zone of the reactor pressure vessel made from steel 15Cr2NiMoVN Class 1 with finished sizes Dext=4.655 mm, Dint=4.240 mm, H=4.910 mm (height for heat treatment – 5.750 mm) is presented. For the first time in Russia in production site of OMZ-Special steel LLC a unique elongated shell blank of the reactor vessel active zone was made from ingot 420.0 t for WWER-TOI project fully meeting the specified requirements in terms of metallurgical quality and set of service properties [ru

Optimization of a membrane reactor for hydrogen production with genetic algorithms

International Nuclear Information System (INIS)

Raceanu, Mircea; Iordache, Ioan; Curuia, Marian; Rasoi, Gabriel; Patularu, Laurentiu; Enache, Adrian

2009-01-01

Full text: Hydrogen is produced via steam reforming of hydrocarbons such as natural gas or methane by using conventional systems. Unfortunately, these systems need at least four different stages, consisting of three reactors and a purification system. Moreover, the steam reforming reaction is an endothermic thermodynamically limited system, meaning that high temperature energy supply is needed for complete conversion. Among different technologies related to production, separation and purification of H 2 , membrane technologies seem to really play a fundamental role. The specific thermodynamic limits are overcome using the so-called membrane reactors, systems in which both reaction and separation occur simultaneously. The hydrogen is driven across the membrane by the pressure difference, depending on the temperature, pressure and reactor length the methane can be completely converted and consequently very pure hydrogen is produced. A membrane reactor has two components which can be optimized namely, the membrane and the reactor dimensions. This paper presents a study on optimization of membrane reactor for enhancing the overall production. A mathematical heterogeneous model of the reactor was used for optimization of reactor performance. Genetic algorithms were used as powerful methods for optimization of complex problems. (authors)

Modelling of the anti-neutrino production and spectra from a Magnox reactor

Science.gov (United States)

Mills, Robert W.; Mountford, David J.; Coleman, Jonathon P.; Metelko, Carl; Murdoch, Matthew; Schnellbach, Yan-Jie

2018-01-01

The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152) during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of monitored reactor

Modelling of the anti-neutrino production and spectra from a Magnox reactor

Directory of Open Access Journals (Sweden)

Mills Robert W

2018-01-01

Full Text Available The anti-neutrino source properties of a fission reactor are governed by the production and beta decay of the radionuclides present and the summation of their individual anti-neutrino spectra. The fission product radionuclide production changes during reactor operation and different fissioning species give rise to different product distributions. It is thus possible to determine some details of reactor operation, such as power, from the anti-neutrino emission to confirm safeguards records. Also according to some published calculations, it may be feasible to observe different anti-neutrino spectra depending on the fissile contents of the reactor fuel and thus determine the reactor's fissile material inventory during operation which could considerable improve safeguards. In mid-2014 the University of Liverpool deployed a prototype anti-neutrino detector at the Wylfa R1 station in Anglesey, United Kingdom based upon plastic scintillator technology developed for the T2K project. The deployment was used to develop the detector electronics and software until the reactor was finally shutdown in December 2015. To support the development of this detector technology for reactor monitoring and to understand its capabilities, the National Nuclear Laboratory modelled this graphite moderated and natural uranium fuelled reactor with existing codes used to support Magnox reactor operations and waste management. The 3D multi-physics code PANTHER was used to determine the individual powers of each fuel element (8×6152 during the year and a half period of monitoring based upon reactor records. The WIMS/TRAIL/FISPIN code route was then used to determine the radionuclide inventory of each nuclide on a daily basis in each element. These nuclide inventories were then used with the BTSPEC code to determine the anti-neutrino spectra and source strength using JEFF-3.1.1 data. Finally the anti-neutrino source from the reactor for each day during the year and a half of

Carbon-14 production in nuclear reactors

International Nuclear Information System (INIS)

Davis, W. Jr.

1977-01-01

The radioactive nuclide 14 C is formed in all nuclear reactors due to absorption of neutrons by carbon, nitrogen, or oxygen. These may be present as components of the fuel, moderator, or structural hardware, or they may be present as impurities. Most of the 14 C formed in the fuels or in the graphite of HTGRs will be converted to a gaseous form at the fuel reprocessing plant, primarily as carbon dioxide; this will be released to the environment unless special equipment is installed to collect it and convert it to a solid for essentially permanent storage. If the 14 C is released as carbon dioxide or in any other chemical form, it will enter the biosphere, be inhaled or ingested as food by nearly all living organisms including man, and will thus contribute to the radiation burden of these organisms. Detailed estimates are presented of the amounts of 14 C formed in LWRs, HTGR, and LMFBR with emphasis on those pathways that are likely to lead to the release of this nuclide, either at the reactor site or at the fuel reprocessing plant. 83 references

36 CFR 223.219 - Sustainable harvest of special forest products.

Science.gov (United States)

2010-07-01

... 36 Parks, Forests, and Public Property 2 2010-07-01 2010-07-01 false Sustainable harvest of....219 Sustainable harvest of special forest products. (a) Sustainable harvest levels. Prior to offering... product's sustainable harvest level. A special forest product's sustainable harvest level is the total...

RA research reactor - potentials and prospective

International Nuclear Information System (INIS)

Sotic, O.

1984-01-01

Since December 1959, the RA reactor was operated successfully, except for a few shorter periods needed for maintenance and a four longer shutdown periods caused by decrease in the heavy water quality. Accordingly, reconstruction of some reactor systems was started at the beginning of this decad, as well as increase of its experimental potential which would enable its efficient reliable operation in the future period. Reconstruction is concerned with emergency core cooling system, special ventilation system, and modernization of the reactor instrumentation. Improvement of the experimental potential is related to modifications of the neutron scattering instruments. Development of methods for isotope production is described as well. Design of the reactor experimental loop with external cooling system will be of significant importance in improvement of reactor potential in the future

Nuclear graphite development, operational problems, and resolution of these problems at the Hanford production reactors

International Nuclear Information System (INIS)

Morgan, W.C.

1996-01-01

This paper chronicles the history of the Hanford Production Reactor, from the initial design considerations for B, D, and F Reactors through the selection of the agreed method for safe disposal of the decommissioned reactors. The operational problems that challenged the operations and support staff of each new generation of production reactors, the engineering actions an operational changes that alleviated or resolved the immediate problems, the changes in reactor design and design-bases for the next generation of production reactors, and the changes in manufacturing variables that resulted in new ''improved'' grades of nuclear graphites for use in the moderators of the Hanford Production Reactors are reviewed in the context of the existing knowledge-base and the mission-driven priorities on the time. 14 refs, 6 figs, 3 tabs

Mechanical design of a magnetic fusion production reactor

International Nuclear Information System (INIS)

Neef, W.S.; Jassby, D.L.

1986-01-01

The mechanical aspects of a tandem mirror and tokamak concepts for the tritium production mission are compared, and a proposed breeding blanket configuration for each type of reactor is presented in detail, along with a design outline of the complete fusion reaction system. In both cases, the reactor design is developed sufficiently to permit preliminary cost estimates of all components. A qualitative comparison is drawn between both concepts from the view of mechanical design and serviceability, and suggestions are made for technology proof tests on unique mechanical features. Detailed cost breakdowns indicate less than 10% difference in the overall costs of the two reactors

New Production Reactor project-management plan

International Nuclear Information System (INIS)

McCrosson, F.J.; Hibbard, L.; Buckner, M.R.

1982-01-01

This document provides a project management plan for the first phase of a project to design and build a new production reactor (NPR) at SRP. The design of the NPR is based upon proven SRP heavy water reactor design, with several enhancements such as full containment, moderator detritiation, improved cooling, and modernized control rooms and instrumentation. The first phase of the NPR project includes environmental and safety analyses, preparation of the technical data summary and basic data, site studies, engineering studies, and conceptual design. The project management plan was developed by a 14-member task force comprised of representatives from the Technical Division, the Manufacturing Division, the Departmental Engineer's Office, and the Engineering Department

Hybrid reactors: Nuclear breeding or energy production?

International Nuclear Information System (INIS)

Piera, Mireia; Lafuente, Antonio; Abanades, Alberto; Martinez-Val, J.M.

2010-01-01

After reviewing the long-standing tradition on hybrid research, an assessment model is presented in order to characterize the hybrid performance under different objectives. In hybrids, neutron multiplication in the subcritical blanket plays a major role, not only for energy production and nuclear breeding, but also for tritium breeding, which is fundamental requirement in fusion-fission hybrids. All three objectives are better achieved with high values of the neutron multiplication factor (k-eff) with the obvious and fundamental limitation that it cannot reach criticality under any event, particularly, in the case of a loss of coolant accident. This limitation will be very important in the selection of the coolant. Some general considerations will be proposed, as guidelines for assessing the hybrid potential in a given scenario. Those guidelines point out that hybrids can be of great interest for the future of nuclear energy in a framework of Sustainable Development, because they can contribute to the efficient exploitation of nuclear fuels, with very high safety features. Additionally, a proposal is presented on a blanket specially suited for fusion-fission hybrids, although this reactor concept is still under review, and new work is needed for identifying the most suitable blanket composition, which can vary depending on the main objective of the hybrid.

An experimental investigation of fission product release in SLOWPOKE-2 reactors - Data report

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

The results of an investigation into the release of fission products from SLOWPOKE-2 reactors fuelled with a highly-enriched uranium alloy core are detailed in Volume 1. This data report (Volume 2) contains plots of the activity concentrations of the fission products observed in the reactor container at the University of Toronto, Ecole Polytechnique and the Kanata Isotope Production Facility. Release rates from the reactor container water to the gas headspace are also included. (author)

Plutonium-239 production rate study using a typical fusion reactor

Energy Technology Data Exchange (ETDEWEB)

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

2008-05-15

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

Simulation of a porous ceramic membrane reactor for hydrogen production

Energy Technology Data Exchange (ETDEWEB)

Yu, W.; Ohmori, T.; Yamamoto, T.; Endo, A.; Nakaiwa, M.; Hayakawa, T. [National Inst. of Advanced Industrial Science and Technology, Tsukuba (Japan); Itoh, N. [National Inst. of Advanced Industrial Science and Technology, Tsukuba (Japan); Utsunomiya Univ. (Japan). Dept. of Applied Chemistry

2005-08-01

A systematic simulation study was performed to investigate the performance of a porous ceramic membrane reactor for hydrogen production by means of methane steam reforming. The results show that the methane conversions much higher than the corresponding equilibrium values can be achieved in the membrane reactor due to the selective removal of products from the reaction zone. The comparison of isothermal and non-isothermal model predictions was made. It was found that the isothermal assumption overestimates the reactor performance and the deviation of calculation results between the two models is subject to the operating conditions. The effects of various process parameters such as the reaction temperature, the reaction side pressure, the feed flow rate and the steam to methane molar feed ratio as well as the sweep gas flow rate and the operation modes, on the behavior of membrane reactor were analyzed and discussed. (author)

Cobalt-60 production in CANDU reactors

International Nuclear Information System (INIS)

Ross, Michel; Lemire, Christian

2002-01-01

CANDU reactors can produce cobalt-60 very efficiently and with an interesting return on investment. This paper discusses what is needed to convert a CANDU reactor into a cobalt-60 producer: what are the different phases, the safety studies required, the physical modifications needed, and what is the minimum involvement of the utility owning the plant. The past ten years of experience of Hydro-Quebec as a cobalt-60 producer will be reviewed, including the management of the risk of both incident and electricity generation loss, and including the benefits for the utility and its personnel. Originally a simple metal used for centuries as a pigment, cobalt-59 today is transformed into cobalt-60, a radioactive element of unprecedented value. Well known in medicine for cancer treatment, cobalt-60 is also used to sterilize a wide range of disposable medical products used in hospitals and to sanitize pharmaceutical and cosmetic products. Cobalt-60 is proving to be a new and effective solution, in the food sector, for preserving harvests and controlling food-borne diseases, or to advantageously replace certain gases and chemical products which are suspected of being harmful or carcinogenic. There are also other applications, such as: hardening of some plastics, treatment of sewage sludge and elimination of harmful insect populations. With a half-life of 5,3 years, cobalt-60 is a metal not found in nature. It is a radioactive isotope produced by exposing stable nuclei of cobalt-59 to neutrons. One of the best places to find such an important neutron source is a nuclear reactor. High energy gamma rays are then emitted during the process of radioactive decay, where cobalt-60 seeks again its stable state

Chemistry of fission product iodine under nuclear reactor accident conditions

International Nuclear Information System (INIS)

Malinauskas, A.P.; Bell, J.T.

1986-01-01

The radioisotopes of iodine are generally acknowledged to be the species whose release into the biosphere as a result of a nuclear reactor accident is of the greatest concern. In the course of its release, the fission product is subjected to differing chemical environments; these can alter the physicochemical form of the fission product and thus modify the manner and extent to which release occurs. Both the chemical environments which are characteristic of reactor accidents and their effect in determining physical and chemical form of fission product iodine have been studied extensively, and are reviewed in this report. 76 refs

Production of Radioisotopes and Radiopharmaceuticals at the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Duong Van Dong; Pham Ngoc Dien; Bui Van Cuong; Mai Phuoc Tho; Nguyen Thi Thu; Vo Thi Cam Hoa

2014-01-01

After reconstruction, the Dalat Nuclear Research Reactor (DNRR) was inaugurated on March 20th, 1984 with the nominal power of 500 kW. Since then the production of radioisotopes and labelled compounds for medical use was started. Up to now, DNRR is still the unique one in Vietnam. The reactor has been operated safely and effectively with the total of about 37,800 hrs (approximately 1,300 hours per year). More than 90% of its operation time and over 80% of its irradiation capacity have been exploited for research and production of radioisotopes. This paper gives an outline of the radioisotope production programme using the DNRR. The production laboratory and facilities including the nuclear reactor with its irradiation positions and characteristics, hot cells, production lines and equipment for the production of Kits for labelling with 99m Tc and for quality control, as well as the production rate are mentioned. The methods used for production of 131 I, 99m Tc, 51 Cr, 32 P, etc. and the procedures for preparation of radiopharmaceuticals are described briefly. Status of utilization of domestic radioisotopes and radiopharmaceuticals in Vietnam is also reported. (author)

Economic Contributions of Radioisotope Production Reactor in Korea

International Nuclear Information System (INIS)

Nam, Ji Hee; Kim, Seung Su; Moon, Kee Whan

2010-01-01

Radioisotopes (RIs) have been used extensively in the fields of industrial, the agricultural, and the medical applications. Especially the deficiency of radioisotopes such as Mo-99 and I-131 in the medical applications recently is becoming the main issue in our society. Radioisotope with the characteristics of public goods in some aspects is mainly playing as the intermediate inputs or goods in the process of the industrial production, with being expected to produce the economic benefits by creating the new demand in the market or enlarging the value added for the related goods and services. In this study, the contribution effects for Korean economy by the construction and operation of the reactor for radioisotope production would be evaluated the effects produced by the activities such as a RI supplies into domestic industry, the RI exports, the neutron transmutation doping services called NTD, and the exports of RI production reactors

Evaluation of enzymatic reactors for large-scale panose production.

Science.gov (United States)

Fernandes, Fabiano A N; Rodrigues, Sueli

2007-07-01

Panose is a trisaccharide constituted by a maltose molecule bonded to a glucose molecule by an alpha-1,6-glycosidic bond. This trisaccharide has potential to be used in the food industry as a noncariogenic sweetener, as the oral flora does not ferment it. Panose can also be considered prebiotic for stimulating the growth of benefic microorganisms, such as lactobacillus and bifidobacteria, and for inhibiting the growth of undesired microorganisms such as E. coli and Salmonella. In this paper, the production of panose by enzymatic synthesis in a batch and a fed-batch reactor was optimized using a mathematical model developed to simulate the process. Results show that optimum production is obtained in a fed-batch process with an optimum production of 11.23 g/l h of panose, which is 51.5% higher than production with batch reactor.

Hydrogen production from high temperature electrolysis and fusion reactor

International Nuclear Information System (INIS)

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

1978-01-01

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

Galacto-oligosaccharide production with immobilized ß-galactosidase in a packed-bed reactor vs. free ß-galactosidase in a batch reactor

NARCIS (Netherlands)

Warmerdam, A.; Benjamins, E.; Leeuw de, T.F.; Broekhuis, T.A.; Boom, R.M.; Janssen, A.E.M.

2014-01-01

We report here that the usage of immobilized enzyme in a continuous packed bed reactor (PBR) can be a good alternative for GOS production instead of the traditional use of free enzyme in a batch reactor. The carbohydrate composition of the product of the PBR with immobilized enzyme was comparable to

Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

2013-10-15

Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

Investigation reactor D-2201 polypropylene production unit using nuclear technique

International Nuclear Information System (INIS)

Wibisono; Sugiharto; Jefri Simanjuntak

2016-01-01

D-2201 reactor is a unit in the polypropylene production process at Pertamina Refinery Unit III Plaju. Reactor with a capacity of 45 kilo liter is not operated in normal operation condition. The validity of liquid level indicator on the unit is doubtful when refers to the production quality. Gamma source of 150 mCi Cobalt-60 and a scintillation detector had been used to scan the outer wall of the reactor to detect the liquid level during operation with a capacity of 40 %. Measurements were made along the reactor walls with 25 mm scan resolution and 5 seconds time sampling. Experiment result shows that the liquid level at the position of 40 % and at normal level position are not observed. Investigation did not find the liquid level above normal. D-2201 is diagnose not normal operating condition diagnosed with liquid abundant passed the recommended limits. Investigation advised to repair or to calibrate the liquid level indicator which is currently installed. (author)

Feasibility of biohydrogen production from cheese whey using a UASB reactor: Links between microbial community and reactor performance

Energy Technology Data Exchange (ETDEWEB)

Castello, E.; Garcia y Santos, C.; Borzacconi, L. [Chemical Engineering Institute, School of Engineering, University of the Republic, Herrera y Reissig 565, Montevideo (Uruguay); Iglesias, T.; Paolino, G.; Wenzel, J.; Etchebehere, C. [Microbiology Department, School of Science and School of Chemistry, University of the Republic, General Flores 2124, Montevideo (Uruguay)

2009-07-15

The present study examines the feasibility of producing hydrogen by dark fermentation using unsterilised cheese whey in a UASB reactor. A lab-scale UASB reactor was operated for more than 250 days and unsterilised whey was used as the feed. The evolution of the microbial community was studied during reactor operation using molecular biology tools (T-RFLP, 16S rRNA cloning library and FISH) and conventional microbiological techniques. The results showed that hydrogen can be produced but in low amounts. For the highest loading rate tested (20 gCOD/L.d), hydrogen production was 122 mL H{sub 2}/L.d. Maintenance of low pH (mean = 5) was insufficient to control methanogenesis; methane was produced concomitantly with hydrogen, suggesting that the methanogenic biomass adapted to the low pH conditions. Increasing the loading rate to values of 2.5 gCOD/gVSS.d favoured hydrogen production in the reactor. Microbiological studies showed the prevalence of fermentative organisms from the genera Megasphaera, Anaerotruncus, Pectinatus and Lactobacillus, which may be responsible for hydrogen production. However, the persistence of methanogenesis and the presence of other fermenters, not clearly recognised as hydrogen producers indicates that competition for the substrate may explain the low hydrogen production. (author)

An experimental investigation of fission product release in SLOWPOKE-2 reactors

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

Long-lived activation products in reactor materials

International Nuclear Information System (INIS)

Evans, J.C.; Lepel, E.L.; Sanders, R.W.; Wilkerson, C.L.; Silker, W.; Thomas, C.W.; Abel, K.H.; Robertson, D.R.

1984-08-01

The purpose of this program was to assess the problems posed to reactor decommissioning by long-lived activation products in reactor construction materials. Samples of stainless steel, vessel steel, concrete, and concrete ingredients were analyzed for up to 52 elements in order to develop a data base of activatable major, minor, and trace elements. Large compositional variations were noted for some elements. Cobalt and niobium concentrations in stainless steel, for example, were found to vary by more than an order of magnitude. A thorough evaluation was made of all possible nuclear reactions that could lead to long lived activation products. It was concluded that all major activation products have been satisfactorily accounted for in decommissioning planning studies completed to date. A detailed series of calculations was carried out using average values of the measured compositions of the appropriate materials to predict the levels of activation products expected in reactor internals, vessel walls, and bioshield materials for PWR and BWR geometries. A comparison is made between calculated activation levels and regulatory guidelines for shallow land disposal according to 10 CFR 61. This analysis shows that PWR and BWR shroud material exceeds the Class C limits and is, therefore, generally unsuitable for near-surface disposal. The PWR core barrel material approaches the Class C limits. Most of the remaining massive components qualify as either Class A or B waste with the bioshield clearly Class A, even at the highest point of activation. Selected samples of activated steel and concrete were subjected to a limited radiochemical analysis program as a verification of the computer model. Reasonably good agreement with the calculations was obtained where comparison was possible. In particular, the presence of 94 Nb in activated stainless steel at or somewhat above expected levels was confirmed

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

Reactor design for minimizing product inhibition during enzymatic lignocellulose hydrolysis II. Quantification of inhibition and suitability of membrane reactors

DEFF Research Database (Denmark)

Andric, Pavle; Meyer, Anne S.; Jensen, Peter Arendt

2010-01-01

conversion are required for alleviation of glucose product inhibition. Supported by numerous calculations this review assesses the quantitative aspects of glucose product inhibition on enzyme-catalyzed cellulose degradation rates. The significance of glucose product inhibition on dimensioning of different......Product inhibition of cellulolytic enzymes affects the efficiency of the biocatalytic conversion of lignocellulosic biomass to ethanol and other valuable products. New strategies that focus on reactor designs encompassing product removal, notably glucose removal, during enzymatic cellulose...... reactor features, including system set-up, dilution rate, glucose output profile, and the problem of cellobiose are examined to illustrate the quantitative significance of the glucose product inhibition and the total glucose concentration on the cellulolytic conversion rate. Comprehensive overviews...

Advanced nuclear fuel production by using fission-fusion hybrid reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

International Nuclear Information System (INIS)

Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

2007-01-01

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered

Charging machine for a fast production reactor

International Nuclear Information System (INIS)

Artem'ev, L.N.; Kurilkin, V.V.

1971-01-01

Charging machine for a fast production reactor is described. The machine contains charging mechanism, mechanism for positioning fresh fuel and spent fuel assemtlies, storage drums with sockets for control rod assemtlies and collet tongs for control rods. Recharging is conducted by means of ramp channel

Fusion reactors for hydrogen production via electrolysis

International Nuclear Information System (INIS)

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

1979-01-01

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

The experimental nuclear reactor: AQUILON; Le reacteur nucleaire experimental: AQUILON

Energy Technology Data Exchange (ETDEWEB)

Girard, Y; Koechlin, J C; Moreau, J M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

'Aquilon' is an experimental reactor specially designed for the neutronic study of heterogeneous multiplying media with solid fuel and liquid moderator. Since this study is in general incompatible with energy production, the power of the reactor has been limited to a minimum so as to be able to obtain a simple and compact structure, easy access, good handling and great flexibility of operation and utilisation. (author) [French] 'Aquilon' est un reacteur experimental specialement concu pour l'etude neutronique de milieux multiplicateurs heterogenes a combustible solide et ralentisseur liquide. Cette etude etant en general incompatible avec la production d'energie, on a limite au minimum la puissance du reacteur pour pouvoir obtenir une structure simple et peu encombrante, un acces facile, une bonne maniabilite et une grande souplesse de fonctionnement et d'utilisation. (auteur)

Production of medical radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for cancer treatment and arterial restenosis therapy after PTCA

International Nuclear Information System (INIS)

Knapp, F.F. Jr.; Beets, A.L.; Mirzadeh, S.; Alexander, C.W.; Hobbs, R.L.

1998-01-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed

New Production Reactors Program Plan

Energy Technology Data Exchange (ETDEWEB)

1990-12-01

Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs.

New Production Reactors Program Plan

International Nuclear Information System (INIS)

1990-12-01

Part I of this New Production Reactors (NPR) Program Plan: describes the policy basis of the NPR Program; describes the mission and objectives of the NPR Program; identifies the requirements that must be met in order to achieve the mission and objectives; and describes and assesses the technology and siting options that were considered, the Program's preferred strategy, and its rationale. The implementation strategy for the New Production Reactors Program has three functions: Linking the design, construction, operation, and maintenance of facilities to policies requirements, and the process for selecting options. The development of an implementation strategy ensures that activities and procedures are consistent with the rationale and analysis underlying the Program. Organization of the Program. The strategy establishes plans, organizational structure, procedures, a budget, and a schedule for carrying out the Program. By doing so, the strategy ensures the clear assignment of responsibility and accountability. Management and monitoring of the Program. Finally, the strategy provides a basis for monitoring the Program so that technological, cost, and scheduling issues can be addressed when they arise as the Program proceeds. Like the rest of the Program Plan, the Implementation Strategy is a living document and will be periodically revised to reflect both progress made in the Program and adjustments in plans and policies as they are made. 21 figs., 5 tabs

Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Werner, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hummel, Andrew J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, John C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, Robert C. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dion, Axel M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wright, Richard N. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ananth, Krishnan P. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-11-01

The Special Purpose Reactor (SPR) is a small 5 MWt, heat pipe-cooled, fast reactor based on the Los Alamos National Laboratory (LANL) Mega-Power concept. The LANL concept features a stainless steel monolithic core structure with drilled channels for UO2 pellet stacks and evaporator sections of the heat pipes. Two alternative active core designs are presented here that replace the monolithic core structure with simpler and easier to manufacture fuel elements. The two new core designs are simply referred to as Design A and Design B. In addition to ease of manufacturability, the fuel elements for both Design A and Design B can be individually fabricated, assembled, inspected, tested, and qualified prior to their installation into the reactor core leading to greater reactor system reliability and safety. Design A fuel elements will require the development of a new hexagonally-shaped UO2 fuel pellet. The Design A configuration will consist of an array of hexagonally-shaped fuel elements with each fuel element having a central heat pipe. This hexagonal fuel element configuration results in four radial gaps or thermal resistances per element. Neither the fuel element development, nor the radial gap issue are deemed to be serious and should not impact an aggressive reactor deployment schedule. Design B uses embedded arrays of heat pipes and fuel pins in a double-wall tank filled with liquid metal sodium. Sodium is used to thermally bond the heat pipes to the fuel pins, but its usage may create reactor transportation and regulatory challenges. An independent panel of U.S. manufacturing experts has preliminarily assessed the three SPR core designs and views Design A as simplest to manufacture. Herein are the results of a preliminary neutronic, thermal, mechanical, material, and manufacturing assessment of both Design A and Design B along with comparisons to the LANL concept (monolithic core structure). Despite the active core differences, all three reactor concepts behave

Fission product poisoning in KS-150 reactor operation

International Nuclear Information System (INIS)

Rana, S.B.

1978-01-01

A three-dimensional model of the KS-150 reactor was used to study reactivity changes induced by reactor poisoning with fission products Xe 135 and Sm 149 . A comparison of transients caused by the poisoning showed the following differences: (1) the duration of the transient Xe poisoning (2 days) is shorter by one order of magnitude than the duration of Sm poisoning (20 days); however, the level of Xe poisoning is greater approximately by one order than the level of the Sm poisoning; (2) the level of steady-state Xe poisoning depends on the output level of the reactor; steady-state Sm poisoning does not depend on this level; (3) following reactor shutdown Xe poisoning may increase to the maximum value of up to Δrhosub(Xe)=20% and will then gradually decrease; Sm poisoning may reach maximum values of up to Δrhosub(Sm)=2% and does not decrease. (J.B.)

From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

Neutronic study of fusion reactor blanket

International Nuclear Information System (INIS)

Barre, F.

1984-02-01

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

Neutronic study of fusion reactor blanket

International Nuclear Information System (INIS)

Barre, F.

1983-06-01

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

Special features of the inverse-beta-decay reaction proceeding on a proton in a reactor-antineutrino flux

Energy Technology Data Exchange (ETDEWEB)

Kopeikin, V. I., E-mail: kopeikin46@yandex.ru; Skorokhvatov, M. D., E-mail: skorokhvatov-md@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

2017-03-15

The evolution of the reactor-antineutrino spectrum and the evolution of the spectrum of positrons from the inverse-beta-decay reaction in the course of reactor operation and after reactor shutdown are considered. The present-day status in determining the initial reactor-antineutrino spectrum on the basis of spectra of beta particles from mixtures of products originating from uranium and plutonium fission is described. A local rise of the experimental spectrum of reactor antineutrinos with respect to the expected spectrum is studied.

Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

International Nuclear Information System (INIS)

Moore, R.S.; Notz, K.J.

1989-06-01

The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

Isotopic evidence for nitrous oxide production pathways in a partial nitritation-anammox reactor.

Science.gov (United States)

Harris, Eliza; Joss, Adriano; Emmenegger, Lukas; Kipf, Marco; Wolf, Benjamin; Mohn, Joachim; Wunderlin, Pascal

2015-10-15

Nitrous oxide (N2O) production pathways in a single stage, continuously fed partial nitritation-anammox reactor were investigated using online isotopic analysis of offgas N2O with quantum cascade laser absorption spectroscopy (QCLAS). N2O emissions increased when reactor operating conditions were not optimal, for example, high dissolved oxygen concentration. SP measurements indicated that the increase in N2O was due to enhanced nitrifier denitrification, generally related to nitrite build-up in the reactor. The results of this study confirm that process control via online N2O monitoring is an ideal method to detect imbalances in reactor operation and regulate aeration, to ensure optimal reactor conditions and minimise N2O emissions. Under normal operating conditions, the N2O isotopic site preference (SP) was much higher than expected - up to 40‰ - which could not be explained within the current understanding of N2O production pathways. Various targeted experiments were conducted to investigate the characteristics of N2O formation in the reactor. The high SP measurements during both normal operating and experimental conditions could potentially be explained by a number of hypotheses: i) unexpectedly strong heterotrophic N2O reduction, ii) unknown inorganic or anammox-associated N2O production pathway, iii) previous underestimation of SP fractionation during N2O production from NH2OH, or strong variations in SP from this pathway depending on reactor conditions. The second hypothesis - an unknown or incompletely characterised production pathway - was most consistent with results, however the other possibilities cannot be discounted. Further experiments are needed to distinguish between these hypotheses and fully resolve N2O production pathways in PN-anammox systems. Copyright © 2015 Elsevier Ltd. All rights reserved.

Study on conceptual design system of tritium production fusion reactor

International Nuclear Information System (INIS)

He Kaihui

2004-11-01

Conceptual design of an advanced tritium production reactor based on spherical torus, which is intermediate application of fusion energy, was presented. Different from traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can within vacuum vessel in order to produce 1 kg excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented. Besides systematical analyses; design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (author)

Fuel performance and fission product behaviour in gas cooled reactors

International Nuclear Information System (INIS)

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

Some in-reactor loop experiments on corrosion product transport and water chemistry

International Nuclear Information System (INIS)

Balakrishnan, P.V.; Allison, G.M.

1978-01-01

A study of the transport of activated corrosion products in the heat transport circuit of pressurized water-cooled nuclear reactors using an in-reactor loop showed that the concentration of particulate and dissolved corrosion products in the high-temperature water depends on such chemical parameters as pH and dissolved hydrogen concentration. Transients in these parameters, as well as in temperature, generally increase the concentration of suspended corrosion products. The maximum concentration of particles observed is much reduced when high-flow, high-temperature filtration is used. Filtration also reduces the steady-state concentration of particles. Dissolved corrosion products are mainly responsible for activity accumulation on surfaces. The data obtained from this study were used to estimate the rate constants for some of the transfer processes involved in the contamination of the primary heat transport circuit in water-cooled nuclear power reactors

Building reactor operator sustain expert system with C language integrated production system

International Nuclear Information System (INIS)

Ouyang Qin; Hu Shouyin; Wang Ruipian

2002-01-01

The development of the reactor operator sustain expert system is introduced, the capability of building reactor operator sustain expert system is discussed with C Language Integrated Production System (Clips), and a simple antitype of expert system is illustrated. The limitation of building reactor operator sustain expert system with Clips is also discussed

Polarized electrode enhances biological direct interspecies electron transfer for methane production in upflow anaerobic bioelectrochemical reactor.

Science.gov (United States)

Feng, Qing; Song, Young-Chae; Yoo, Kyuseon; Kuppanan, Nanthakumar; Subudhi, Sanjukta; Lal, Banwari

2018-08-01

The influence of polarized electrodes on the methane production, which depends on the sludge concentration, was investigated in upflow anaerobic bioelectrochemical (UABE) reactor. When the polarized electrode was placed in the bottom zone with a high sludge concentration, the methane production was 5.34 L/L.d, which was 53% higher than upflow anaerobic sludge blanket (UASB) reactor. However, the methane production was reduced to 4.34 L/L.d by placing the electrode in the upper zone of the UABE reactor with lower sludge concentration. In the UABE reactor, the methane production was mainly improved by the enhanced biological direct interspecies electron transfer (bDIET) pathway, and the methane production via the electrode was a minor fraction of less than 4% of total methane production. The polarized electrodes that placed in the bottom zone with a high sludge concentration enhance the bDIET for methane production in the UABE reactor and greatly improve the methane production. Copyright © 2018. Published by Elsevier Ltd.

Analysis of reactor-neutrino spectra fully based on gross theory of beta-decay emphasizing the special role of odd-odd FP nuclides

Science.gov (United States)

Yoshida, Tadashi; Tachibana, Takahiro; Chiba, Satoshi

2017-09-01

We applied the gross theory of β-decay to calculate the reactor antineutrino (νe) spectra emitted from 235,238U and 239,241Pu samples under neutron irradiation by summing up all the contributions from a large number of decaying fission-products (FPs). Considering the special role of the odd(Z)-odd(N) FPs in spectrum-shaping, we utilized the experimentally-known spin-parity of each odd-odd FP through the treatment proposed by Nakata, Tachibana and Yamada. Owing to this treatment, the consistency between calculated and experimental spectra was remarkably improved in a way expected from the nature of the gross theory of β-decay.

Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

International Nuclear Information System (INIS)

Chi-Mei Lee; Kuo-Tsang Hung

2006-01-01

The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

International Nuclear Information System (INIS)

Fehrenbach, P.J.; Wood, J.C.

1987-06-01

Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

Research reactor production of radioisotopes for medical use

International Nuclear Information System (INIS)

Mani, R.S.

1985-01-01

More than 70% of all radioisotopes applied in medical diagnosis and research are currently produced in research reactors. Research reactors are also an important source of certain radioisotopes, such as 60 Co, 90 Y, 137 Cs and 198 Au, which are employed in teletherapy and brachytherapy. For regular medical applications, mainly 29 radionuclides produced in research reactors are used. These are now produced on an 'industrial scale' by many leading commercial manufacturers in industrialized countries as well as by national atomic energy establishments in developing countries. Five main neutron-induced reactions have been employed for the regular production of these radionuclides, namely: (n,γ), (n,p), (n,α), (n,γ) followed by decay, and (n, fission). In addition, the Szilard-Chalmers process has been used in low- and medium-flux research reactors to enrich the specific activity of a few radionuclides (mainly 51 Cr) produced by the (n,γ) reaction. Extensive work done over the last three decades has resulted in the development of reliable and economic large-scale production methods for most of these radioisotopes and in the establishment of rigorous specifications and purity criteria for their manifold applications in medicine. A useful spectrum of other radionuclides with suitable half-lives and low to medium toxicity can be produced in research reactors, with the requisite purity and specific activity and at a reasonable cost, to be used as tracers. Thanks to the systematic work done in recent years by many radiopharmaceutical scientists, the radionuclides of several elements, such as arsenic, selenium, rhenium, ruthenium, palladium, cadmium, tellurium, antimony, platinum, lead and the rare earth elements, which until recently were considered 'exotic' in the biomedical field, are now gaining attention. (author)

Hydrogen production by high temperature electrolysis of water vapour and nuclear reactors

International Nuclear Information System (INIS)

Jean-Pierre Py; Alain Capitaine

2006-01-01

This paper presents hydrogen production by a nuclear reactor (High Temperature Reactor, HTR or Pressurized Water Reactor, PWR) coupled to a High Temperature Electrolyser (HTE) plant. With respect to the coupling of a HTR with a HTE plant, EDF and AREVA NP had previously selected a combined cycle HTR scheme to convert the reactor heat into electricity. In that case, the steam required for the electrolyser plant is provided either directly from the steam turbine cycle or from a heat exchanger connected with such cycle. Hydrogen efficiency production is valued using high temperature electrolysis. Electrolysis production of hydrogen can be performed with significantly higher thermal efficiencies by operating in the steam phase than in the water phase. The electrolysis performance is assessed with solid oxide and solid proton electrolysis cells. The efficiency from the three operating conditions (endo-thermal, auto-thermal and thermo-neutral) of a high temperature electrolysis process is evaluated. The technical difficulties to use the gases enthalpy to heat the water are analyzed, taking into account efficiency and technological challenges. EDF and AREVA NP have performed an analysis to select an optimized process giving consideration to plant efficiency, plant operation, investment and production costs. The paper provides pathways and identifies R and D actions to reach hydrogen production costs competitive with those of other hydrogen production processes. (authors)

A two-stage ethanol-based biodiesel production in a packed bed reactor

DEFF Research Database (Denmark)

Xu, Yuan; Nordblad, Mathias; Woodley, John

2012-01-01

were conducted in a simulated series of reactors by repeatedly passing the reaction mixture through a single reactor, with separation of the by-product glycerol and water between passes in the first and second stages, respectively. The second stage brought the major components of biodiesel to ‘in......-spec’ levels according to the European biodiesel specifications for methanol-based biodiesel. The highest overall productivity achieved in the first stage was 2.52 kg FAEE(kg catalyst)−1 h−1 at a superficial velocity of 7.6 cm min−1, close to the efficiency of a stirred tank reactor under similar conditions...

The operation characteristics of biohydrogen production in continuous stirred tank reactor with molasses

Energy Technology Data Exchange (ETDEWEB)

Hong, C.; Wei, H.; Jie-xuan, D.; Xin, Y.; Chuan-ping, Y. [Northeast Forestry Univ., Harbin (China). School of Forestry; Li, Y.F. [Northeast Forestry Univ., Harbin (China). School of Forestry; Shanghai Univ. Engineering, Shanghai (China). College of Chemistry and Chemical Engineering

2010-07-01

The anaerobic fermentation biohydrogen production in a continuous stirred tank reactor (CSTR) was investigated as a means for treating molasses wastewater. The research demonstrated that the reactor has the capacity of continuously producing hydrogen in an initial biomass (as volatile suspension solids) of 17.74 g/L, temperature of approximately 35 degrees Celsius, hydraulic retention time of 6 hours. The reactor could begin the ethanol-type fermentation in 12 days and realize stable hydrogen production. The study also showed that the CSTR reactor has a favourable stability even with an organic shock loading. The hydrogen yield and chemical oxygen demand (COD) increased, as did the hydrogen content.

Specificities of micro-structured reactors for hydrogen production and purification

Energy Technology Data Exchange (ETDEWEB)

Dupont, N.; Germani, G.; Van Veen, A.C.; Schuurman, Y.; Mirodatos, C. [Institut de Recherches sur la Catalyse - CNRS, 2, Avenue Albert Einstein, 69626 Villeurbanne Cedex (France); Schaefer, G. [Atotech Deutschland GmbH, PO Box 210780, 10507 Berlin (Germany)

2007-07-15

This paper presents the specificities of micro-structured reactors as compared to conventional fixed-bed reactors through two case studies devoted to (i) hydrogen production by methanol steam reforming, (ii) hydrogen purification by water-gas shift (WGS). Key features like catalyst coating stability, temperature and pressure management, effects of operating conditions (residence time, pressure drops, etc.) are well identified as controlling the micro-reactor performances for methanol reforming. These devices are also shown to be excellent tools for fast access to reaction kinetics as exemplified for the WGS reaction, subject to operating conditions carefully chosen to ensure proper hydrodynamics, in order to use conventional plug flow reactor models for extracting rate constants. (author)

Production of high-specific activity radionuclides using SM high-flux reactor

International Nuclear Information System (INIS)

Karelin, Ye.A.; Toporov, Yu.G.; Filimonov, V.T.; Vakhetov, F.Z.; Tarasov, V.A.; Kuznetsov, R.A.; Lebedev, V.M.; Andreev, O.I.; Melnik, M.I.; Gavrilov, V.D.

1997-01-01

The development of High Specific Activity (HSA) radionuclides production technologies is one of the directions of RIAR activity, and the high flux research reactor SM, having neutron flux density up to 2.10 15 cm -2 s 1 in a wide range of neutron spectra hardness, plays the principal role in this development. The use of a high-flux reactor for radionuclide production provides the following advantages: production of radionuclides with extremely high specific activity, decreasing of impurities content in irradiated targets (both radioactive and non-radioactive), cost-effective use of expensive isotopically enriched target materials. The production technologies of P-33, Gd-153, W-188, Ni-63, Fe-55,59, Sn-113,117m,119m, Sr- 89, applied in industry, nuclear medicine, research, etc, were developed by RIAR during the last 5-10 years. The research work included the development of calculation procedures for radionuclide reactor accumulation forecast, experimental determination of neutron cross-sections, the development of irradiated materials reprocessing procedures, isolation and purification of radionuclides. The principal results are reviewed in the paper. (authors)

Effect of ionite decomposition products on the reactor coolant pH in a boiling-water reactor

International Nuclear Information System (INIS)

Bredikhin, V.Ya.; Moskvin, L.N.

1982-01-01

The effect of products resulting from thermal radiolysis of ionites on water-chemical regime of NPP with RBMK is considered basing on investigations conducted in a boiling type experimental reactor. Data are presented on dynamics of changes in the specific electric conductivity and pH of the coolant following destruction of ion exchange groups and ionite matrix under the effect of reactor radiation. The authors draw a conclusion that radiation destruction of ionito fine disperse suspension or high-molecular soluble compounds in the reactor are, probably, one of the main reasons for variations in pH values of the coolant at NPP in non-correction water chemical regime

Plasma-arc reactor for production possibility of powdered nano-size materials

International Nuclear Information System (INIS)

Hadzhiyski, V; Mihovsky, M; Gavrilova, R

2011-01-01

Nano-size materials of various chemical compositions find increasing application in life nowadays due to some of their unique properties. Plasma technologies are widely used in the production of a range of powdered nano-size materials (metals, alloys, oxides, nitrides, carbides, borides, carbonitrides, etc.), that have relatively high melting temperatures. Until recently, the so-called RF-plasma generated in induction plasma torches was most frequently applied. The subject of this paper is the developments of a new type of plasma-arc reactor, operated with transferred arc system for production of disperse nano-size materials. The new characteristics of the PLASMALAB reactor are the method of feeding the charge, plasma arc control and anode design. The disperse charge is fed by a charge feeding system operating on gravity principle through a hollow cathode of an arc plasma torch situated along the axis of a water-cooled wall vertical tubular reactor. The powdered material is brought into the zone of a plasma space generated by the DC rotating transferred plasma arc. The arc is subjected to Auto-Electro-Magnetic Rotation (AEMR) by an inductor serially connected to the anode circuit. The anode is in the form of a water-cooled copper ring. It is mounted concentrically within the cylindrical reactor, with its lower part electrically insulated from it. The electric parameters of the arc in the reactor and the quantity of processed charge are maintained at a level permitting generation of a volumetric plasma discharge. This mode enables one to attain high mean mass temperature while the processed disperse material flows along the reactor axis through the plasma zone where the main physico-chemical processes take place. The product obtained leaves the reactor through the annular anode, from where it enters a cooling chamber for fixing the produced nano-structure. Experiments for AlN synthesis from aluminium power and nitrogen were carried out using the plasma reactor

Transformation products of clindamycin in moving bed biofilm reactor (MBBR)

DEFF Research Database (Denmark)

Ooi, Gordon Tze Hoong; Escola Casas, Monica; Andersen, Henrik Rasmus

2017-01-01

Clindamycin is widely prescribed for its ability to treat a number of common bacterial infections. Thus, clindamycin enters wastewater via human excretion or disposal of unused medication and widespread detection of pharmaceuticals in rivers proves the insufficiency of conventional wastewater...... treatment plants in removing clindamycin. Recently, it has been discovered that attached biofilm reactors, e.g., moving bed biofilm reactors (MBBRs) obtain a higher removal of pharmaceuticals than conventional sludge wastewater treatment plants. Therefore, this study investigated the capability of MBBRs...... process converts clindamycin into the, possibly persistent, products clindamycin sulfoxide and N-desmethyl clindamycin as well as 3 other mono-oxygenated products. Subsequently, the removal kinetics of clindamycin and the formation of the two identified products were investigated in batch experiments...

Fuel performance and fission product behaviour in gas cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-11-01

The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

Molt salts reactors capacity for wastes incineration and energy production

International Nuclear Information System (INIS)

David, S.; Nuttin, A.

2005-01-01

The molten salt reactors present many advantages in the framework of the IV generation systems development for the energy production and/or the wastes incineration. After a recall of the main studies realized on the molten salt reactors, this document presents the new concepts and the identified research axis: the MSRE project and experience, the incinerators concepts, the thorium cycle. (A.L.B.)

Radioisotope Production Plan and Strategy of Kijang Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Kye Hong; Lee, Jun Sig [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

This reactor will be located at Kijang, Busan, Korea and be dedicated to produce mainly medical radioisotopes. Tc-99m is very important isotope for diagnosis and more than 80% of radiation diagnostic procedures in nuclear medicine depend on this isotope. There were, however, several times of insecure production of Mo-99 due to the shutdown of major production reactors worldwide. OECD/NEA is leading member countries to resolve the shortage of this isotope and trying to secure the international market of Mo-99. The radioisotope plan and strategy of Kijang Research Reactor (KJRR) should be carefully established to fit not only the domestic but also international demand on Mo-99. The implementation strategy of 6 principles of HLG-MR should be established that is appropriate to national environments. Ministry of Science, ICT and Future Planning and Ministry of Health and welfare should cooperate well to organize the national radioisotope supply structure, to set up the reasonable and competitive pricing of radioisotopes, and to cope with the international supply strategy.

Isotopic alloying to tailor helium production rates in mixed spectrum reactors

International Nuclear Information System (INIS)

Mansur, L.K.; Rowcliffe, A.F.; Grossbeck, M.L.; Stoller, R.E.

1985-01-01

The purposes of this work are to increase the understanding of mechanisms by which helium affects microstructure and properties, to aid in the development of materials for fusion reactors, and to obtain data from fission reactors in regimes of direct interest for fusion reactor applications. Isotopic alloying is examined as a means of manipulating the ratio of helium transmutations to atom displacements in mixed spectrum reactors. The application explored is based on artificially altering the relative abundances of the stable isotopes of nickel to systematically vary the fraction of 58 Ni in nickel bearing alloys. The method of calculating helium production rates is described. Results of example calculations for proposed experiments in the High Flux Isotope Reactor are discussed

Reactors Save Energy, Costs for Hydrogen Production

Science.gov (United States)

2014-01-01

While examining fuel-reforming technology for fuel cells onboard aircraft, Glenn Research Center partnered with Garrettsville, Ohio-based Catacel Corporation through the Glenn Alliance Technology Exchange program and a Space Act Agreement. Catacel developed a stackable structural reactor that is now employed for commercial hydrogen production and results in energy savings of about 20 percent.

Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

2014-05-01

The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

Investigation of special capsule technologies for material in-pile irradiation test and development plan in HANARO

Energy Technology Data Exchange (ETDEWEB)

Cho, M. S.; Son, J. M.; Kim, D. S.; Park, S. J.; Cho, Y. G.; Seo, C. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

2002-10-01

In-pile test for several materials such as Zr alloy, stainless steel, Cr-Ni steel etc. which are used as structural material of the advanced reactor and KNGR(Korea Next Generation Reactor) like SMART, is necessary to produce the design data for developing new reactor materials. Advanced countries like USA, Europe and Japan etc. are not only performing the simple irradiation test for materials, but developing many kinds of special capsule to perform in-pile test having special purpose. For the special test items of fuel rod, fission products, total heat generation, swelling, deformation, sweep gas, temperature ramping and BOCA etc. are being actively concerned. There are capsules measuring creep, fatigue, crack growth, and controlling fluence etc. for special irradiation test of materials. In addition, the advanced countries are developing several instrument technologies suitable for the special capsules. In HANARO, non-instrumented, instrumented material capsules and non-instrumented fuel capsule have been developed and they have been utilized in the irradiation test for users, and creep capsule loading single specimen was made and is planned to test in the reactor soon. For some forthcoming years, special capsules not only measuring creep deformation with multi-specimens, fatigue, controlling fluence but crack propagation and gas sweep considering the requirements of users will be developed in HANARO.

Achievements of European projects on membrane reactor for hydrogen production

NARCIS (Netherlands)

di Marcoberardino, G.; Binotti, M.; Manzolini, G.; Viviente, J.L.; Arratibel Plazaola, A.; Roses, L.; Gallucci, F.

2017-01-01

Membrane reactors for hydrogen production can increase both the hydrogen production efficiency at small scale and the electric efficiency in micro-cogeneration systems when coupled with Polymeric Electrolyte Membrane fuel cells. This paper discusses the achievements of three European projects

Removal of spent fuel from the TVR reactor for reprocessing and proposals for the RA reactor spent fuel handling

International Nuclear Information System (INIS)

Volkov, E.B.; Konev, V.N.; Shvedov, O.V.; Bulkin, S.Yu; Sokolov, A.V.

2002-01-01

The 2,5 MW heavy-water moderated and cooled research reactor TVR was located at the Moscow Institute for Theoretical and Experimental Physics site. In 1990 the final batch of spent nuclear fuel (SNF) from the TVR reactor was transported for reprocessing to Production Association (PA) 'Mayak'. This transportation of the SNF was a part of TVR reactor decommissioning. The special technology and equipment was developed in order to fulfill the preparation of TVR SNF for transportation. The design of the TVR reactor and the fuel elements used are similar to the design and fuel elements of the RA reactor. Two different ways of RA spent fuel elements for transportation to reprocessing plant are considered: in aluminum barrels, and in additional cans. The experience and equipment used for the preparing TVR fuel elements for transportation can help the staff of RA reactor to find the optimal way for these technical operations. (author)

Argon-41 production and evolution at the Oregon State University TRIGA Reactor (OSTR)

International Nuclear Information System (INIS)

Anellis, L.G.; Johnson, A.G.; Higginbotham, J.F.

1988-01-01

In this study, argon-41 concentrations were measured at various locations within the reactor facility to assess the accuracy of models used to predict argon-41 evolution from the reactor tank, and to determine the relationship between argon gas evolution from the tank and subsequent argon-41 concentrations throughout the reactor room. In particular, argon-41 was measured directly above the reactor tank with the reactor tank lids closed, at other accessible locations on the reactor top with the tank lids both closed and open, and at several locations on the first floor of the reactor room. These measured concentrations were then compared to values calculated using a modified argon-41 production and evolution model for TRIGA reactor tanks and ventilation values applicable to the OSTR facility. The modified model was based in part on earlier TRIGA models for argon-41 production and release, but added features which improved the agreement between predicted and measured values. The approximate dose equivalent rate due to the presence of argon-41 in reactor room air was calculated for several different locations inside the OSTR facility. These dose rates were determined using the argon-41 concentration measured at each specific location, and were subsequently converted to a predicted quarterly dose equivalent for each location based on the reactor's operating history. The predicted quarterly dose equivalent values were then compared to quarterly doses measured by film badges deployed as dose-integrating area radiation monitors at the locations of interest. The results indicate that the modified production and evolution model is able to predict argon-41 concentrations to within a factor of ten when compared to the measured data. Quarterly dose equivalents calculated from the measured argon-41 concentrations and the reactor's operating history seemed consistent with results obtained from the integrating area radiation monitors. Given the argon-41 concentrations measured

USE OF THE MODULAR HELIUM REACTOR FOR HYDROGEN PRODUCTION

International Nuclear Information System (INIS)

SCHULTZ, K.R.

2003-01-01

OAK-B135 A significant ''Hydrogen Economy'' is predicted that will reduce our dependence on petroleum imports and reduce pollution and greenhouse gas emissions. Hydrogen is an environmentally attractive fuel that has the potential to displace fossil fuels, but contemporary hydrogen production is primarily based on fossil fuels. The author has recently completed a three-year project for the US Department of Energy (DOE) whose objective was to ''define an economically feasible concept for production of hydrogen, using an advanced high-temperature nuclear reactor as the energy source''. Thermochemical water-slitting, a chemical process that accomplishes the decomposition of water into hydrogen and oxygen, met this objective. The goal of the first phase of this study was to evaluate thermochemical processes which offer the potential for efficient, cost-effective, large-scale production of hydrogen, and to select one for further detailed consideration. They selected the Sulfur-Iodine cycle. In the second phase, they reviewed all the basic reactor types for suitability to provide the high temperature heat needed by the selected thermochemical water splitting cycle and chose the helium gas-cooled reactor. In the third phase they designed the chemical flowsheet for the thermochemical process and estimated the efficiency and cost of the process and the projected cost of producing hydrogen. These results are summarized in this report

Current utilization of research reactor on radioisotopes production in China

International Nuclear Information System (INIS)

Liu Yishu

2000-01-01

The main technical parameters of the four research reactors and their current utilization status in radioisotope manufacture and labeling compounds preparation are described. The radioisotopes, such as Co-60 sealed source, Ir-192 sealed source, γ-knife source, I-131, I-125, Sm-153, P-32 series products, In-113m generator, Tc-99m gel generator, Re-188 gel generator, C-14, Ba-131, Sr-89, 90 Y, etc., and their labeling compounds prepared from the reactor produced radionuclides, such as I-131-MIBG, I-131-Hippure, I-131-capsul, Sm-153-EDTMP, Re-186-HEDP, Re-186-HA, C-14-urea, and radioimmunoassay kits etc. are presented as well. Future development plan of radioisotopes and labeling compounds in China is also given. Simultaneously, the possibility and methods of bilateral or multilateral co-operation in utilization of research reactor, personnel and technology exchange of radioisotope production and labeling compounds is also discussed. (author)

Current utilization of research reactor on radioisotopes production in China

Energy Technology Data Exchange (ETDEWEB)

Liu Yishu [Nuclear Power Institute of China, Chengdu (China)

2000-10-01

The main technical parameters of the four research reactors and their current utilization status in radioisotope manufacture and labeling compounds preparation are described. The radioisotopes, such as Co-60 sealed source, Ir-192 sealed source, {gamma}-knife source, I-131, I-125, Sm-153, P-32 series products, In-113m generator, Tc-99m gel generator, Re-188 gel generator, C-14, Ba-131, Sr-89, {sup 90}Y, etc., and their labeling compounds prepared from the reactor produced radionuclides, such as I-131-MIBG, I-131-Hippure, I-131-capsul, Sm-153-EDTMP, Re-186-HEDP, Re-186-HA, C-14-urea, and radioimmunoassay kits etc. are presented as well. Future development plan of radioisotopes and labeling compounds in China is also given. Simultaneously, the possibility and methods of bilateral or multilateral co-operation in utilization of research reactor, personnel and technology exchange of radioisotope production and labeling compounds is also discussed. (author)

Transient fission-product release during reactor shutdown and startup

International Nuclear Information System (INIS)

Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

1997-12-01

Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

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

Environmental aspects of fusion reactors

International Nuclear Information System (INIS)

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

1975-01-01

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

Fission-product burnup chain model for research reactor application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jung Do; Gil, Choong Sup; Lee, Jong Tai [Korea Atomic Energy Research Inst., Daeduk (Republic of Korea)

1990-12-01

A new fission-product burnup chain model was developed for use in research reactor analysis capable of predicting the burnup-dependent reactivity with high precision over a wide range of burnup. The new model consists of 63 nuclides treated explicitly and one fissile-independent pseudo-element. The effective absorption cross sections for the preudo-element and the preudo-element yield of actinide nuclides were evaluated in the this report. The model is capable of predicting the high burnup behavior of low-enriched uranium-fueled research reactors.(Author).

Transient fission product release during reactor shutdown and startup

International Nuclear Information System (INIS)

Hunt, C.E.L.; Lewis, B.J.

1995-01-01

Sweep gas experiments performed at CRL from 1979 to 1985 have been analysed to determine the fraction of the fission product gas inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the xenon release from companion fuel elements and from a well documented experimental fuel bundle irradiated in the NRU reactor. The measured gas release could be matched to within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. (author)

Commercial Aspect of Research Reactor Fuel Element Production

International Nuclear Information System (INIS)

Susanto, B.G; Suripto, A

1998-01-01

Several aspects affecting the commercialization of the Research Reactor Fuel Element Production Installation (RR FEPI) under a BUMN (state-owned company)have been studied. The break event point (BEP) value based on total production cost used is greatly depending upon the unit selling price of the fuel element. At a selling price of USD 43,500/fuel element, the results of analysis shows that the BEP will be reached at 51% of minimum available capacity. At a selling price of US$ 43.500/fuel element the total income (after tax) for 7 years ahead is US $ 4.620.191,- The net present value in this study has a positive value is equal to US $ 2.827.527,- the internal rate of return will be 18% which is higher than normal the bank interest rare (in US dollar) at this time. It is concluded therefore that the nuclear research reactor fuel element produced by state-owned company BUMN has a good prospect to be sold commercially

Conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production fusion reactor based on spherical torus, which is intermediate application of fusion energy, was presented in this paper. Differing from the traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and maximize tritium breeding ratio with arrangement of tritium production blankets within vacuum vessel as possible in order to produce 1 kg excess tritium except need of self-sufficient plasma core with 40% or more corresponding plant availability. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented, providing the backgrounds and reference for next detailed conceptual design

Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

International Nuclear Information System (INIS)

Popa-Simil, Liviu

2008-01-01

Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages - possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools. (author)

General scheme of research reactor mainly for production of fission 99Mo

International Nuclear Information System (INIS)

Shen Feng; Liu Xingmin; Wu Xiaochun; Sun Zheng; Guo Chunqiu; Yi Dayong

2012-01-01

On the basis of the analysis for current circumstance and development tendency of research reactor mainly for 99 Mo production in the world, the design idea of this sort of research reactor was proposed. By the optimization and basic design, the general design parameters of the reactor were analyzed and testified. The evaluation of output activities of 99 Mo and the analysis of economics were conducted on the basically assumption. It is argued that the economics of this reactor is improved dramatically while the safety is ensured by the analysis. (authors)

Utilization of the Dalat Research Reactor for Radioisotope Production, Neutron Activation Analysis, Research and Training

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Duong Van Dong; Cao Dong Vu; Nguyen Xuan Hai

2013-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool type reactor loaded with a mixed core of HEU (36% enrichment) and LEU (19.75% enrichment) fuel assemblies. The reactor is used as a neutron source for the purposes of radioisotopes production, neutron activation analysis, basic and applied research and training. The reactor is operated mainly in continuous runs of 108 hours for cycles of 3–4 weeks for the above mentioned purposes. The current status of safety, operation and utilization of the reactor is given and some aspects for improvement of commercial products and services of the DNRR are also discussed in this paper. (author)

RA Research reactor, Annual report 1972

International Nuclear Information System (INIS)

Milosevic, D. et al.

1972-12-01

During 1972, the total production was 31151 MWh which is 3.8% higher than planned. The reactor was used for irradiation and experiments according to the demand of 381 users, of which 340 from the Institute and 41 external users. This report contains detailed data about reactor power and experiments performed in 1972. Discrepancies from the action plan, meaning higher production was achieved due to special demands of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to announced power cuts. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. There were no longer interruptions. One shorter interruption (shorter than 24 hours) caused by removal of a UO 2 capsule from the core, placed there for measuring heat transfer. Total personnel exposure dose was lower than during previous years. One accident caused contamination with gases and aerosols containing mainly shot-living isotopes. Decontamination od surfaces was less than during previous years. Practically there was no surface contamination that would demand action of the decontamination team, except for the regular decontamination after refueling. It was concluded that the successful operation in 1972 has a special significance having taking in account the financial crisis caused by the unresolved status of the reactor. It is emphasised, in the plan for the next year that there is an urgent need of making a long-term plan of rector application. It is indispensable to finish preparatory tasks for replacing the fuel with the highly enriched fuel elements by 1974, and building the core emergency cooling system [sr

Radioactive Contamination Near Natural Uranium - Graphite - Gas Reactors

International Nuclear Information System (INIS)

Chassany, J.; Pouthier, J.

1967-01-01

The authors give the results of numerous assessments of contamination in connection with reactors in operation during maintenance; reactors shut down during overhaul and repair work (coolants, exchangers, interior of the tank, etc.) ; and accidents in the cooling circuit and ruptured cladding. They show that, except in special cases, it is mainly activation products that predominate. Moreover, after eight years of operation the points where contamination likely to give considerable dose rates accumulates remain very localized, and there has been no need to reinforce personnel protection measures. (author) [fr

Corrosion products, activity transport and deposition in boiling water reactor recirculation systems

International Nuclear Information System (INIS)

Alder, H.P.; Buckley, D.; Grauer, R.; Wiedemann, K.H.

1992-01-01

The deposition of activated corrosion products in the recirculation loops of Boiling Water Reactors produces increased radiation levels which lead to a corresponding increase in personnel radiation dose during shut down and maintenance. The major part of this dose rate is due to cobalt-60. Based on a comprehensive literature study concerning this theme, it has been attempted to identify the individual stages of the activity build-up and to classify their importance. The following areas are discussed in detail: The origins of the corrosion products and of cobalt-59 in the reactor feedwaters; the consolidation of the cobalt in the fuel pins deposits (activation); the release and transport of cobalt-60; the build-up of cobalt-60 in the corrosion products in the recirculation loops. Existing models of the build-up of circuit radioactivity are discussed and the operating experiences from selected reactors are summarized. 90 refs, figs and tabs

Butanol production by Clostridium acetobutylicum in a continuous packed bed reactor.

Science.gov (United States)

Napoli, Fabio; Olivieri, Giuseppe; Russo, Maria Elena; Marzocchella, Antonio; Salatino, Piero

2010-06-01

In this study, we report on a butanol production process by immobilized Clostridium acetobutylicum in a continuous packed bed reactor (PBR) using Tygon rings as a carrier. The medium was a solution of lactose (15-30 g/L) and yeast extract (3 g/L) to emulate the cheese whey, an abundant lactose-rich wastewater. The reactor was operated under controlled conditions with respect to the pH and to the dilution rate. The pH and the dilution rate ranged between 4 and 5, the dilution rate between 0.54 and 2.4 h(-1) (2.5 times the maximum specific growth rate assessed for suspended cells). The optimal performance of the reactor was recorded at a dilution rate of 0.97 h(-1): the butanol productivity was 4.4 g/Lh and the selectivity of solvent in butanol was 88%(w).

Operational safety and reactor life improvements of Kyoto University Reactor

International Nuclear Information System (INIS)

Utsuro, M.; Fujita, Y.; Nishihara, H.

1990-01-01

Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

1999-01-01

The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

An experimental study of the behaviour of fission products following an accident on a swimming pool reactor

International Nuclear Information System (INIS)

Dadillon, J.

1976-11-01

In the estimation of nuclear risks connected with the running of a reactor an essential factor, sometimes neglected because insufficiently known, is the knowledge of the type, amount and behaviour of the contamination actually released inside the containment in the case of an accident. In the special case of swimming pool reactors the cooling fluid proves to be a very efficient barrier against contamination. Three experiments were carried out in the reactor CABRI, during which several fuel element plates were melted inside the core itself. (Author)

Continuous thermophilic biohydrogen production in packed bed reactor

International Nuclear Information System (INIS)

Roy, Shantonu; Vishnuvardhan, M.; Das, Debabrata

2014-01-01

Highlights: • Continuous H 2 production in whole cell immobilized system was compared with CSTR. • Suitability of environment friendly support matrix for immobilization of whole cells was explored. • Pack bed reactor showed higher stability as compared to CSTR at lower HRTs. • Flow cytometry study showed the influence of recycle ratio on viability of cells. • Novel approach to find out the effect of NADH/NAD + ratio during H 2 production. - Abstract: The present research work deals with the performance of packed bed reactor for continuous H 2 production using cane molasses as a carbon source. Maximum H 2 production rate of 1.7 L L −1 h −1 was observed at a dilution rate and recycle ratio of 0.8 h −1 and 0.6, respectively which was corresponding to the lowest NADH/NAD + ratio. This suggests that the utilization of NADH pool for H 2 and metabolite production might lead to decrement in NADH/NAD + ratio. Thus NADH/NAD + ratio show inverse relation with hydrogen production. The substrate degradation kinetics was investigated as a function of flow rate considering the external film diffusion model. At a flow rate of 245 mL h −1 , the contribution of external film mass transfer coefficient and first order substrate degradation constant were 55.4% and 44.6% respectively. Recycle ratio of 0.6 improved the hydrogen production rates by 9%. The viable cell count was directly proportional to the recycle ratio (within the range 0.1–0.6). Taguchi design showed the significant influence of the feed pH on continuous H 2 production followed by dilution rate and recycle ratio. Thus environmentally friendly and cheaper solid matrix like coconut coir could be efficiently used for thermophilic continuous hydrogen production

Status of and prospects for gas-cooled reactors

International Nuclear Information System (INIS)

1984-01-01

The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

Reactor physics challenges in GEN-IV reactor design

Energy Technology Data Exchange (ETDEWEB)

Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

2005-02-15

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

Reactor physics challenges in GEN-IV reactor design

International Nuclear Information System (INIS)

Driscoll, Michael K.; Hejzlar, Pavel

2005-01-01

An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

Development of a novel integrated continuous reactor system for biocatalytic production of biodiesel.

Science.gov (United States)

Chattopadhyay, Soham; Sen, Ramkrishna

2013-11-01

A novel integrated immobilized enzyme-reactor system involving a continuous stirred tank reactor with two packed bed reactors in series was developed for the continuous production of biodiesel. The problem of methanol solubility into oil was solved by introducing a stirred tank reactor to dissolve methanol into partially converted oil. This step made the process perfectly continuous without requiring any organic solvent and intermittent methanol addition in the process. The substrate feeding rate of 0.74 mL/min and enzyme loading of 0.75 g per reactor were determined to be optimum for maximum biodiesel yield. The integrated continuous process was stable up to 45 cycles with biodiesel productivity of 137.2 g/L/h, which was approximately 5 times higher than solvent free batch process. In comparison with the processes reported in literature using expensive Novozyme 435 and hazardous organic solvent, the present process is completely green and perfectly continuous with economic and environmental advantages. Copyright © 2013 Elsevier Ltd. All rights reserved.

Cost analysis of enzymatic biodiesel production in small-scaled packed-bed reactors

NARCIS (Netherlands)

Budzaki, S.; Miljic, G.; Sundaram, S.; Tisma, M.; Hessel, V.

2017-01-01

A cost analysis of enzymatic biodiesel production in small-scaled packed-bed reactors using refined sunflower oil is performed in this work. A few enzymatic micro-flow reactors have so far reached a performance close to gram-scale, which might be sufficient for the pharmaceutical industry. This

Sustainability management for operating organizations of research reactors

Energy Technology Data Exchange (ETDEWEB)

Kibrit, Eduardo; Aquino, Afonso Rodrigues de, E-mail: ekibrit@ipen.br, E-mail: araquino@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

Sustainability management for operating organizations of research reactors

International Nuclear Information System (INIS)

Kibrit, Eduardo; Aquino, Afonso Rodrigues de

2017-01-01

Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

LOFC fission product release and circulating activity calculations for gas-cooled reactors

International Nuclear Information System (INIS)

Apperson, C.E. Jr.; Carruthers, L.M.; Lee, C.E.

1977-01-01

The inventories of fission products in a gas-cooled reactor under accident and normal steady state conditions are time and temperature dependent. To obtain a reasonable estimate of these inventories it is necessary to consider fuel failure, a temperature dependent variable, and radioactive decay, a time dependent variable. Using arbitrary radioactive decay chains and published fuel failure models for the High Temperature Gas-Cooled Reactor (HTGR), methods have been developed to evaluate the release of fission products during the Loss of Forced Circulation (LOFC) accident and the circulating and plateout fission product inventories during steady state non-accident operation. The LARC-2 model presented here neglects the time delays in the release from the HTGR due to diffusion of fission products from particles in the fuel rod through the graphite matrix. It also neglects the adsorption and evaporation process of metallics at the fuel rod-graphite and graphite-coolant hole interfaces. Any time delay due to the finite time of transport of fission products by convection through the coolant to the outside of the prestressed concrete reactor vessel (PCRV) is also neglected. This model assumes that all fission products released from fuel particles are immediately deposited outside the PCRV with no time delay

An Explanatory Study of Lean Practices in Job Shop Production/ Special Job Production/ Discrete Production/ Batch Shop Production Industries

OpenAIRE

Lavlesh Kumar Sharma; Ravindra Mohan Saxena

2014-01-01

In this paper, the study explores the benefits and advantages of Lean Practices or Lean Thinking in Job shop production/ Special job production/ Discrete production/ Batch shop production industries. The Lean Practices have been applied more compatible in Job shop production than in the continuous/ mass production because of several barriers and hurdles in the industrial context that influence the whole processes again and again, this happens due to the lack of knowledge about...

List of the reports from reactor safety research of the BMFT and USNRC

International Nuclear Information System (INIS)

1977-02-01

This list presents a survey of reports from the FRG and USA issued on special problems of reactor safety research. The problems include emergency core cooling, the containment during loss of coolant, core meltdown, component safety, external impacts, risk and reliability, quality assurance, fission product transport, radiation exposure, as well as the LWR, HTR and the fast sodium-cooled reactor in general. (orig./HK) [de

Inspection of the UO2 special fuel for the prototype heavy water reactor 'FUGEN'

International Nuclear Information System (INIS)

Miura, Makoto; Ohmori, Takuro; Yoshino, Hiroyuki; Matsui, Hiromasa; Hirosawa, Naonori

1979-01-01

UO 2 special fuel assemblies are the fuel for material irradiation incorporating irradiation specimens, for the prototype heavy water reactor ''FUGEN''. In order to monitor the behavior of the pressure tube material irradiated with neutrons for long time, monitoring specimens were equipped in the core. This special fuel was fabricated by the Nuclear Fuel Industries, Ltd. (NFI), and the fuel cladding tubes, the capsule guide tubes and the capsule tubes were furnished by PNC. The irradiation specimens were prepared by PNC, and incorporated into the assemblies by NFI. The inspection by PNC on the special fuel assemblies was conducted following the inspection by the maker, which was made on UO 2 pellets, fuel element and assembly parts except cladding tubes, after completing the fabrication. The specifications of the special fuel, especially for the outer and inner layer pellets, the outer and inner layer fuel elements and the fuel assemblies, are presented. The flow sheet for the inspection process and surveillance test of special fuel assemblies is illustrated. The inspection items, the materials and the quantity inspection are tabulated for the fuel elements, the fuel assemblies and the irradiation capsules, respectively. The structure of a special type fuel assembly is shown. For each inspection, the inspection methods and items and the results are explained. As for the results of inspection of the special fuel, the UO 2 pellets, fuel element parts, fuel elements, fuel assembly parts, fuel assemblies, capsules and irradiation specimens were in accordance with the specifications. Regarding the situation of the quality control in the processes, check was made with many documents, and it was recognized that the quality control was performed in the quality assurance program. (Nakai, Y.)

Biodiesel Reactor Design with Glycerol Separation to Increase Biodiesel Production Yield

Directory of Open Access Journals (Sweden)

Budy Rahmat

2013-09-01

Full Text Available The study consisted of reactor design used for transesterification process, effect of glycerol separation ontransesterification reaction, determination of biodiesel quality, and mass balance analysis. The reactor was designed byintegrating circulated pump/stirrer, static mixer, and sprayer that intensify the reaction in the outer tank reactor. The objective was to reduce the use of methanol in excess and to shorten the processing time. The results showed that thereactor that applied the glycerol separation was able to compensate for the decreased use of the reactant methanol from 6:1 to 5:1 molar ratio, and changed the mass balance in the product, including: (i the increase of biodiesel productionfrom 42.37% to 49.34%, and (ii the reduction of methanol in excess from 42.37% to 32.89%. The results suggested that the efficiency of biodiesel production could be increased with the glycerol separation engineering.

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

International Nuclear Information System (INIS)

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

1981-01-01

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

A proposed standard on medical isotope production in fission reactors

International Nuclear Information System (INIS)

Schenter, R. E.; Brown, G. J.; Holden, C. S.

2006-01-01

Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

Computationally-generated nuclear forensic characteristics of early production reactors with an emphasis on sensitivity and uncertainty

International Nuclear Information System (INIS)

Redd, Evan M.; Sjoden, Glenn; Erickson, Anna

2017-01-01

Highlights: •X-10 reactor is used as a case study for nuclear forensic signatures. •S/U analysis is conducted to derive statistically relevant markers. •Computationally-generated signatures aid with proliferation pathway identification. •Highest uncertainty in total plutonium production originates from 238 Pu and 242 Pu. -- Abstract: With nuclear technology and analysis advancements, site access restrictions, and ban on nuclear testing, computationally-generated nuclear forensic signatures are becoming more important in gaining knowledge to a reclusive country’s weapon material production capabilities. In particular, graphite-moderated reactors provide an appropriate case study for isotopics relevant in Pu production in a clandestine nuclear program due to the ease of design and low thermal output. We study the production characteristics of the X-10 reactor with a goal to develop statistically-relevant nuclear forensic signatures from early Pu production. In X-10 reactor, a flat flux gradient and low burnup produce exceptionally pure Pu as evident by the 240 Pu/ 239 Pu ratio. However, these design aspects also make determining reactor zone attribution, done with the 242 Pu/ 240 Pu ratio, uncertain. Alternatively, the same ratios produce statistically differentiable results between Manhattan Project and post-Manhattan Project reactor configurations, allowing for attribution conclusions.

Fission product model for lattice calculation of high conversion boiling water reactor

International Nuclear Information System (INIS)

Iijima, S.; Yoshida, T.; Yamamoto, T.

1988-01-01

A high precision fission product model for boiling water reactor (BWR) lattice calculation was developed, which consists of 45 nuclides to be treated explicitly and one nonsaturating pseudo nuclide. This model is applied to a high conversion BWR lattice calculation code. From a study based on a three-energy-group calculation of fission product poisoning due to full fission products and explicitly treated nuclides, the multigroup capture cross sections and the effective fission yields of the pseudo nuclide are determined, which do not depend on fuel types or reactor operating conditions for a good approximation. Apart from nuclear data uncertainties, the model and the derived pseudo nuclide constants would predict the fission product reactivity within an error of 0.1% Δk at high burnup

Traceability: a demand of agro industrial chain for special products

Directory of Open Access Journals (Sweden)

José Verissimo Foggiatto Silveira

2007-10-01

Full Text Available The inclusion of agricultural products with different nutritional features has altered the relationship, the upstream and the downstream of enterprises that produce and commercialize them. Coordination in the Agro Industrial System is demanded, including traceability as a way to guarantee the conformity of products, attending external clients and agricultural industries that require quality certification. This quality tool enables the identification of some details in the productive chain, such as seeds, farming, harvesting, storage, transportation and industrialization of products. Thus, this essay describes the concept of traceability and provides information of special products from a cooperative from Paraná, which has controlled process in the productive chain, demanded by contractual partnerships done with enterprises that provide fertilizers and food processors. It was identified that this cooperative commercializes three products that need traceability: two special kinds of corn and the regular kind of soybean.

Study on improvement of continuous hydrogen production by photosynthetic biofilm in interior illuminant reactor.

Science.gov (United States)

Liu, Wenhui; Yuan, Linjiang; Wei, Bo

2016-09-01

In the present study, a new type of interior optical fiber illuminating reactor was developed for H2 production to solve the problem of luminous intensity attenuation at the center portion of a reactor, and an immobilization technique was used to enhance the stability of a continuous hydrogen production process with attached photosynthetic bacteria, using glucose as a sole carbon substrate for the indigenous photosynthetic bacteria (PSB) Rhodopseudomonas palustris SP-6. Results of the experiments showed that the interior optical fiber illuminating reactor produces H2 more efficiently and productively than the exterior light source reactor, with the cumulative H2 production, the maximum H2 production rate and H2 yield increased by 813ml, 11.3ml l-1 h-1 and 22.3%, respectively. The stability of the product of continuous hydrogen was realized by immobilizing PSB on the surface of powder active carbon(PAC). After adding the dosage of 2.0g l-1 PAC, the continuous steady operation of H2 production gave a high H2 yield of 1.398 mol H2 mol-1 glucose and an average H2 production rate of 35.1ml l-1 h-1 illuminating with a single interior optical fiber light source. Meanwhile, a higher H2 yield of 1.495 mol H2 mol-1 glucose and an average H2 production rate of 38.7ml l-1 h-1 were attained illuminating with a compound lamp in the continuous H2 production for 20 days.

Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

International Nuclear Information System (INIS)

Bultman, J.

1992-04-01

Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1987-01-01

RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

International Nuclear Information System (INIS)

Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

2009-01-01

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor

International Nuclear Information System (INIS)

Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

2015-01-01

In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe

A novel approach to the production of medical radioisotopes: the homogeneous SLOWPOKE reactor

Energy Technology Data Exchange (ETDEWEB)

Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [retired, Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Royal Canadian Navy, Ottawa, Ontario (Canada)

2015-03-15

In 2009, the unexpected 15-month outage of the Canadian NRU nuclear reactor resulted in a sudden 30% world shortage, with higher shortages experienced in North America than in Europe. Commercial radioisotope production is from just eight nuclear reactors, most being aging systems near the end of their service life. This paper proposes a more efficient production and distribution model. Tc-99m unit doses would be distributed to regional hospitals from ten integrated 'industrial radiopharmacies', located at existing licensed nuclear reactor sites in North America. At each site, one or more 20 kW Homogeneous SLOWPOKE nuclear reactors would deliver 15 litres of irradiated aqueous uranyl sulfate fuel solution daily to industrial-scale hot cells, for extraction of Mo-99; and the low-enriched uranium would be recycled. Purified Mo-99 would be incorporated in large Mo-99/Tc-99m generators for extraction of Tc-99m five days a week; and each automated hot-cell facility would be designed to load up to 7,000 Tc-99m syringes daily for road delivery to all of the nuclear medicine hospitals within a 3-hour range. At the current price of $20 per unit dose, the annual gross income from 10 sites would be approximately $360 million. The Homogeneous SLOWPOKE reactor evolved from the inherently safe SLOWPOKE-2 research reactor, with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors at the end-of-core life, enabling them to continue their primary missions of research and education, together with full time commercial radioisotope production. The Homogeneous SLOWPOKE reactor was modelled using both deterministic and probabilistic reactor simulation codes. The homogeneous fuel mixture is a dilute aqueous solution of low-enriched uranyl sulfate containing approximately 1 kg of U-235. The reactor is controlled by mechanical absorber rods in the beryllium reflector. Safety analysis was carried out for both normal operation and transient conditions. The most severe

Modelling and operation of reactors for enzymatic biodiesel production

DEFF Research Database (Denmark)

Price, Jason Anthony

to the production of high fructose corn syrup, upgrading of fats and oils and biodiesel production to name a few. Despite these examples of industrial enzymatic applications, it is still not “clear cut” how to implement biocatalyst in industry and how best to optimize the processes. This is because the processing...... aspects of the enzyme with reaction/reactor engineering is performed. This strategy is applied to a case study of biodiesel production catalysed by a liquid enzyme formulation. The use of enzymes for biodiesel production is still in its infancy with non-optimized process designs. Furthermore is it unclear...

Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1985-07-01

The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

International Nuclear Information System (INIS)

1985-01-01

The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

Considerations in the design of a high power medical isotope production reactor

International Nuclear Information System (INIS)

Ball, Russell M.; Nordyke, William H.; Brown, Roy

2002-01-01

For the low enriched aqueous homogeneous reactor to be economic in the production of medical isotopes, such as Mo-99 and Sr-89, the power level should be of the order of 100 kWth. This is double the earlier designs and this paper discusses the design changes which must be considered to meet this goal. The topics considered are: 1. Heat removal from the reactor solution; 2. Recombination of radiolytic gases; 3. Adequate radiation shielding; 4. Stability of reactor power with fluctuating reactivity; 5. Adequate cooling of the reflector; 6. Independent shutdown mechanisms; 7. Required volume of the reactor; 8. Economic implementation. (author)

Chemical immobilization of fission products reactive with nuclear reactor components

International Nuclear Information System (INIS)

Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

1975-01-01

This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

Air quality impact analysis in support of the new production reactor environmental impact statement

International Nuclear Information System (INIS)

Hadley, D.L.

1991-04-01

The Pacific Northwest Laboratory (PNL) conducted this air quality impact analysis for the US Department of Energy (DOE). The purpose of this work was to provide Argonne National Laboratory (ANL) with the required estimates of ground-level concentrations of five criteria air pollutants at the Hanford Site boundary from each of the stationary sources associated with the new production reactor (NPR) and its supporting facilities. The DOE proposes to provide new production capacity for the primary production of tritium and secondary production of plutonium to support the US nuclear weapons program. Three alternative reactor technologies are being considered by DOE: the light-water reactor, the low-temperature, heavy-water reactor, and the modular high-temperature, gas-cooled reactor. In this study, PNL provided estimates of the impacts of the proposed action on the ground-level concentration of the criteria air pollutants for each of the alternative technologies. The criteria pollutants were sulfur dioxide, nitrogen dioxide, carbon monoxide, total suspended particulates, and particulates with a diameter of less than 10 microns. Ground-level concentrations were estimated for the peak construction phase activities expected to occur in 1997 and for the operational phase activities beginning in the year 2000. Ground-level concentrations of the primary air pollutants were estimated to be well below any of the applicable national or state ambient air quality standards. 12 refs., 19 tabs

Method of fission product beta spectra measurements for predicting reactor anti-neutrino emission

Energy Technology Data Exchange (ETDEWEB)

Asner, D.M.; Burns, K.; Campbell, L.W.; Greenfield, B.; Kos, M.S., E-mail: markskos@gmail.com; Orrell, J.L.; Schram, M.; VanDevender, B.; Wood, L.S.; Wootan, D.W.

2015-03-11

The nuclear fission process that occurs in the core of nuclear reactors results in unstable, neutron-rich fission products that subsequently beta decay and emit electron antineutrinos. These reactor neutrinos have served neutrino physics research from the initial discovery of the neutrino to today's precision measurements of neutrino mixing angles. The prediction of the absolute flux and energy spectrum of the emitted reactor neutrinos hinges upon a series of seminal papers based on measurements performed in the 1970s and 1980s. The steadily improving reactor neutrino measurement techniques and recent reconsiderations of the agreement between the predicted and observed reactor neutrino flux motivates revisiting the underlying beta spectra measurements. A method is proposed to use an accelerator proton beam delivered to an engineered target to yield a neutron field tailored to reproduce the neutron energy spectrum present in the core of an operating nuclear reactor. Foils of the primary reactor fissionable isotopes placed in this tailored neutron flux will ultimately emit beta particles from the resultant fission products. Measurement of these beta particles in a time projection chamber with a perpendicular magnetic field provides a distinctive set of systematic considerations for comparison to the original seminal beta spectra measurements. Ancillary measurements such as gamma-ray emission and post-irradiation radiochemical analysis will further constrain the absolute normalization of beta emissions per fission. The requirements for unfolding the beta spectra measured with this method into a predicted reactor neutrino spectrum are explored.

Radioactive effluent sources and special system of channels on the RA Reactor at the Boris Kidric Institute; Izvori radioaktivnih efluenata i specijalna kanalizacija u reaktoru RA Instituta Boris Kidric

Energy Technology Data Exchange (ETDEWEB)

Bojovic, P; Gacinovic, O; Milosevic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

1964-10-15

The paper describes the place of origin, composition and activity of radioactive effluents appearing in some reactor systems and special channels for carrying these effluents to disposal basins located outside the reactor building (author)

Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

Science.gov (United States)

Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

2017-08-01

The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

Needs and accuracy requirements for fission product nuclear data in the physics design of power reactor cores

International Nuclear Information System (INIS)

Rowlands, J.L.

1978-01-01

The fission product nuclear data accuracy requirements for fast and thermal reactor core performance predictions were reviewed by Tyror at the Bologna FPND Meeting. The status of the data was assessed at the Meeting and it was concluded that the requirements of thermal reactors were largely met, and the yield data requirements of fast reactors, but not the cross section requirements, were met. However, the World Request List for Nuclear Data (WRENDA) contains a number of requests for fission product capture cross sections in the energy range of interest for thermal reactors. Recent reports indicate that the fast reactor reactivity requirements might have been met by integral measurements made in zero power critical assemblies. However, there are requests for the differential cross sections of the individual isotopes to be determined in addition to the integral data requirements. The fast reactor requirements are reviewed, taking into account some more recent studies of the effects of fission products. The sodium void reactivity effect depends on the fission product cross sections in a different way to the fission product reactivity effect in a normal core. This requirement might call for different types of measurement. There is currently an interest in high burnup fuel cycles and alternative fuel cycles. These might require more accurate fission product data, data for individual isotopes and data for capture products. Recent calculations of the time dependence of fission product reactivity effects show that this is dependent upon the data set used and there are significant uncertainties. Some recent thermal reactor studies on approximations in the treatment of decay chains and the importance of xenon and samarium poisoning are also summarized. (author)

The different facilities of the reactor Phenix for radio isotope production and fission product burner

International Nuclear Information System (INIS)

Coulon, P.; Clerc, R.; Tommasi, J.

1993-01-01

During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (authors). 9 figs

Research and development program of hydrogen production system with high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Miyamoto, Y.; Shiozawa, S.; Ogawa, M.; Inagaki, Y.; Nishihara, T.; Shimizu, S.

2000-01-01

Japan Atomic Energy Research Institute (JAERI) has been developing a hydrogen production system with a high temperature gas-cooled reactor (HTGR). While the HTGR hydrogen production system has the following advantages compared with a fossil-fired hydrogen production system; low operation cost (economical fuel cost), low CO 2 emission and saving of fossil fuel by use of nuclear heat, it requires some items to be solved as follows; cost reduction of facility such as a reactor, coolant circulation system and so on, development of control and safety technologies. As for the control and safety technologies, JAERI plans demonstration test with hydrogen production system by steam reforming of methane coupling to 30 Wt HTGR, named high temperature engineering test reactor (HTTR). Prior to the demonstration test, a 1/30-scale out-of-pile test facility is in construction for safety review and detailed design of the HTTR hydrogen production system. Also, design study will start for reduction of facility cost. Moreover, basic study on hydrogen production process without CO 2 emission is in progress by thermochemical water splitting. (orig.)

Power generation versus fuel production in light water hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1977-06-01

The economic potentials of fissile-fuel-producing light-water hybrid reactors (FFP-LWHR) and of fuel-self-sufficient (FSS) LWHR's are compared. A simple economic model is constructed that gives the capital investment allowed for the hybrid reactor so that the cost of electricity generated in the hybrid based energy system equals the cost of electricity generated in LWR's. The power systems considered are LWR, FSS-LWHR, and FFP-LWHR plus LWR, both with and without plutonium recycling. The economic potential of FFP-LWHR's is found superior to that of FSS-LWHR's. Moreover, LWHR's may compete, economically, with LWR's. Criteria for determining the more economical approach to hybrid fuel or power production are derived for blankets having a linear dependence between F and M. The examples considered favor the power generation rather than fuel production

Temperature oscillations in methanol partial oxidation reactor for the production of hydrogen

Energy Technology Data Exchange (ETDEWEB)

Kim, Jinsu; Byeon, Jeonguk; Seo, Il Gyu; Lee, Hyun Chan; Kim, Dong Hyun; Lee, Jietae [Kyungpook National University, Daegu (Korea, Republic of)

2013-04-15

Methanol partial oxidation (POX) is a well-known reforming reaction for the production of hydrogen from methanol. Since POX is relatively fast and highly exothermic, this reforming method will be efficient for the fast start-up and load-following operation. However, POX generates hot spots around catalyst and even oscillations in the reactor temperature. These should be relieved for longer operations of the reactor without catalyst degradations. For this, temperature oscillations in a POX reactor are investigated experimentally. Various patterns of temperature oscillations according to feed flow rates of reactants and reactor temperatures are obtained. The bifurcation phenomena from regular oscillations to chaotic oscillations are found as the methanol flow rate increases. These experimental results can be used for theoretical analyses of oscillations and for designing safe reforming reactors.

Temperature oscillations in methanol partial oxidation reactor for the production of hydrogen

International Nuclear Information System (INIS)

Kim, Jinsu; Byeon, Jeonguk; Seo, Il Gyu; Lee, Hyun Chan; Kim, Dong Hyun; Lee, Jietae

2013-01-01

Methanol partial oxidation (POX) is a well-known reforming reaction for the production of hydrogen from methanol. Since POX is relatively fast and highly exothermic, this reforming method will be efficient for the fast start-up and load-following operation. However, POX generates hot spots around catalyst and even oscillations in the reactor temperature. These should be relieved for longer operations of the reactor without catalyst degradations. For this, temperature oscillations in a POX reactor are investigated experimentally. Various patterns of temperature oscillations according to feed flow rates of reactants and reactor temperatures are obtained. The bifurcation phenomena from regular oscillations to chaotic oscillations are found as the methanol flow rate increases. These experimental results can be used for theoretical analyses of oscillations and for designing safe reforming reactors

Comparison of actinides and fission products recycling scheme with the normal plutonium recycling scheme in fast reactors

Directory of Open Access Journals (Sweden)

Salahuddin Asif

2013-01-01

Full Text Available Multiple recycling of actinides and non-volatile fission products in fast reactors through the dry re-fabrication/reprocessing atomics international reduction oxidation process has been studied as a possible way to reduce the long-term potential hazard of nuclear waste compared to that resulting from reprocessing in a wet PUREX process. Calculations have been made to compare the actinides and fission products recycling scheme with the normal plutonium recycling scheme in a fast reactor. For this purpose, the Karlsruhe version of isotope generation and depletion code, KORIGEN, has been modified accordingly. An entirely novel fission product yields library for fast reactors has been created which has replaced the old KORIGEN fission products library. For the purposes of this study, the standard 26 groups data set, KFKINR, developed at Forschungszentrum Karlsruhe, Germany, has been extended by the addition of the cross-sections of 13 important actinides and 68 most important fission products. It has been confirmed that these 68 fission products constitute about 95% of the total fission products yield and about 99.5% of the total absorption due to fission products in fast reactors. The amount of fissile material required to guarantee the criticality of the reactor during recycling schemes has also been investigated. Cumulative high active waste per ton of initial heavy metal is also calculated. Results show that the recycling of actinides and fission products in fast reactors through the atomics international reduction oxidation process results in a reduction of the potential hazard of radioactive waste.

Renewal of reactor cooling system of JMTR. Reactor building site

International Nuclear Information System (INIS)

Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

2012-03-01

The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

Basic study on high temperature gas cooled reactor technology for hydrogen production

International Nuclear Information System (INIS)

Chang, Jong Hwa; Lee, W. J.; Lee, H. M.

2003-01-01

The annual production of hydrogen in the world is about 500 billion m 3 . Currently hydrogen is consumed mainly in chemical industries. However hydrogen has huge potential to be consumed in transportation sector in coming decades. Assuming that 10% of fossil energy in transportation sector is substituted by hydrogen in 2020, the hydrogen in the sector will exceed current hydrogen consumption by more than 2.5 times. Currently hydrogen is mainly produced by steam reforming of natural gas. Steam reforming process is chiefest way to produce hydrogen for mass production. In the future, hydrogen has to be produced in a way to minimize CO2 emission during its production process as well as to satisfy economic competition. One of the alternatives to produce hydrogen under such criteria is using heat source of high-temperature gas-cooled reactor. The high-temperature gas-cooled reactor represents one type of the next generation of nuclear reactors for safe and reliable operation as well as for efficient and economic generation of energy

Nuclear reactors and fuel cycle

International Nuclear Information System (INIS)

2014-01-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

Nuclear reactors and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

Advanced Reactors Around the World

International Nuclear Information System (INIS)

Majumdar, Debu

2003-01-01

At the end of 2002, 441 nuclear power plants were operating around the globe and providing 17% of the world's electricity. Although the rate of population growth has slowed, recent United Nations data suggest that two billion more people will be added to the world by 2050. A special report commissioned by the Intergovernmental Panel on Climate Change estimated that electricity demand would grow almost eight-fold from 2000 to 2050 in a high economic grown scenario and more than double in a low-growth scenario. There is also a global aspiration to keep the environment pristine. Because of these reasons, it is expected that a large number of new nuclear reactors may be operating by 2050. Realization of this has created an impetus for the development of a new generation of reactors in several countries. The goal is to make nuclear power cost-competitive with other resources and to enhance safety to a level that no evacuation outside a plant site would be necessary. It should also generate less waste, prevent materials diversion for weapons production, and be sustainable. This article discusses the status of next-generation reactors under development around the world. Specifically highlighted are efforts related to the Generation IV International Forum (GIF) and its six reactor concepts for research and development: Very High Temperature Reactor (VHTR); Gas-Cooled Fast Reactor (GFR); Supercritical Water-Cooled Reactor (SCWR); Sodium-Cooled Fast Reactor (SFR); Lead-Cooled Fast Reactor (LFR); and Molten Salt Reactor (MSR). Also highlighted are nuclear activities specific to Russia and India

Use of LEU in the aqueous homogeneous medical isotope production reactor

Energy Technology Data Exchange (ETDEWEB)

Ball, R.M. [Babock & Wilcox, Lynchburg, VA (United States)

1997-08-01

The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution.

Use of LEU in the aqueous homogeneous medical isotope production reactor

International Nuclear Information System (INIS)

Ball, R.M.

1997-01-01

The Medical Isotope Production Reactor (MIPR) is an aqueous solution of uranyl nitrate in water, contained in an aluminum cylinder immersed in a large pool of water which can provide both shielding and a medium for heat exchange. The control rods are inserted at the top through re-entrant thimbles. Provision is made to remove radiolytic gases and recombine emitted hydrogen and oxygen. Small quantities of the solution can be continuously extracted and replaced after passing through selective ion exchange columns, which are used to extract the desired products (fission products), e.g. molybdenum-99. This reactor type is known for its large negative temperature coefficient, the small amount of fuel required for criticality, and the ease of control. Calculation using TWODANT show that a 20% U-235 enriched system, water reflected can be critical with 73 liters of solution

Membrane steam reforming of natural gas for hydrogen production by utilization of medium temperature nuclear reactor

International Nuclear Information System (INIS)

Djati Hoesen Salimy

2010-01-01

The assessment of steam reforming process with membrane reactor for hydrogen production by utilizing of medium temperature nuclear reactor has been carried out. Difference with the conventional process of natural gas steam reforming that operates at high temperature (800-1000°C), the process with membrane reactor operates at lower temperature (~500°C). This condition is possible because the use of perm-selective membrane that separate product simultaneously in reactor, drive the optimum conversion at the lower temperature. Besides that, membrane reactor also acts the role of separation unit, so the plant will be more compact. From the point of nuclear heat utilization, the low temperature of process opens the chance of medium temperature nuclear reactor utilization as heat source. Couple the medium temperature nuclear reactor with the process give the advantage from the point of saving fossil fuel that give direct implication of decreasing green house gas emission. (author)

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

International Nuclear Information System (INIS)

Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F.

2009-01-01

The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U 235 (typically Pu 242 , Np 237 , U 238 , Th 232 ). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

Analysis of an homogeneous solution reactor for 99 Mo production

International Nuclear Information System (INIS)

Weir, A.; Lopasso, E.; Gho, C.

2007-01-01

The 99m Tc is the more used radioisotope in nuclear medicine, used in 80% of procedures of nuclear medicine in the world. This is due to their characteristics practically ideal for the diagnostic. The 99m Tc is obtained by decay of the 99 Mo, which can produce it by irradiating enriched targets in 98 Mo, or as fission product, irradiating uranium targets or by means of homogeneous solution reactors. The pattern of the used reactor in the neutron analysis possesses a liquid fuel composed of uranyl nitrate dissolved in water with the attach of nitric acid. This solution is contained in a cylindrical recipient of stainless steel reflected with light water. The reactor is refrigerated by means of an helicoidal heat exchanger immersed in the fuel solution. The heat of the fuel is removed by natural convection while the circulation of the water inside the exchanger is forced. The control system of the reactor consists on 6 independent cadmium bars, with followers of water. An auxiliary control system can be the level of the fuel solution inside container tank, but it was not included in the pattern in study. One studies the variations of the reactivity of the system due to different phenomena. An important factor during the normal operation of the reactor is the variation of temperature taking to a volumetric expansion of the fuel and ghastly effects in the same one. Another causing phenomenon of changes in the reactivity is the variation of the concentration of uranium in the combustible solution. An important phenomenon in this type of reactors is the hole fraction in the nucleus I liquidate due to the radiolysis and the possible boil of the water of the combustible solution. Some of the possible cases of abnormal operation were studied as the lost one of coolant in the secondary circuit of the heat exchanger, the introduction and evaporation of water in the nucleus. The reactivity variations were studied using the codes of I calculate MCNP, WIMS and TORT. All the

Telemanipulation - a special activity in remotely controlled operations

International Nuclear Information System (INIS)

Rose, K.W.; Andre, Y.

1992-01-01

Work to be done in areas hostile to humans needs special and careful preparation. If short-term entry is possible, groups of men can be trained to do the necessary work. If not, special devices have to be designed, built, and tested on mockups before the real work can be executed. Based on experience gained from maintenance in car production and test programs for a reprocessing facility, it was decided to train a special group of men to do remotely controlled work in hostile areas without endangering them and to use their personal experience as the basis for future work. This is the old-fashioned way of all professions. Some needs to be able to do that remotely controlled work with normally existing operational means and combinations of them like cranes, mechanical and electromechanical master slave manipulators (MMSMs and EMSMs), saws, files, hammer, tig-welding equipment, etc., in air as well as underwater. This paper discusses use of a remote operator manipulator (ROM), remote operator welder (ROW), a test of underwater work, and the repair of two activated jets pumps of a boiling water reactor BWR with a fueling machine, reactor crane, EMSM, and conventional tools

High-temperature nuclear reactor power plant cycle for hydrogen and electricity production – numerical analysis

Directory of Open Access Journals (Sweden)

Dudek Michał

2016-01-01

Full Text Available High temperature gas-cooled nuclear reactor (called HTR or HTGR for both electricity generation and hydrogen production is analysed. The HTR reactor because of the relatively high temperature of coolant could be combined with a steam or gas turbine, as well as with the system for heat delivery for high-temperature hydrogen production. However, the current development of HTR’s allows us to consider achievable working temperature up to 750°C. Due to this fact, industrial-scale hydrogen production using copper-chlorine (Cu-Cl thermochemical cycle is considered and compared with high-temperature electrolysis. Presented calculations show and confirm the potential of HTR’s as a future solution for hydrogen production without CO2 emission. Furthermore, integration of a hightemperature nuclear reactor with a combined cycle for electricity and hydrogen production may reach very high efficiency and could possibly lead to a significant decrease of hydrogen production costs.

Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

International Nuclear Information System (INIS)

Dhami, P.S; Yadav, J.S; Agarwal, K.

2017-01-01

Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

Radionuclide inventory and source terms for the surplus production reactors at Hanford

International Nuclear Information System (INIS)

Miller, R.L.; Steffes, J.M.

1987-01-01

Radionuclide inventories have been estimated for the eight surplus production reactors at Hanford. The inventories listed represent more than 95% of the total curie burden; the remaining 5% is distributed in piping, tunnels, and various other locations within the reactor building and unaccounted for inventories within the reactors or fuel storage basins. Estimates are conservative as the methodology was designed to overestimate the radionuclide inventories in the facilities. The estimated inventory per reactor facility ranges from 13,000 curies to 58,000 curies. The majority of the present inventory consists of tritium, carbon-14, cobalt-60, and nickel-63. The information in this document combines data from past characterization efforts and introduces adjustments for added information and refinement. The inventory of hazardous materials in the reactor facilities is also addressed. This document has been revised to include new reduced inventory figures for chlorine-36. The new figures were derived from recent analysis of irradiated graphite from the 105-kW reactor

Multiscale hydrodynamic investigation to intensify the biogas production in upflow anaerobic reactors.

Science.gov (United States)

Jiang, Jiankai; Wu, Jing; Zhang, Jinbai; Poncin, Souhila; Li, Huai Z

2014-03-01

Hydrodynamics plays a main role for the performance of an anaerobic reactor involving three phases: wastewater, sludge granules and biogas bubbles. The present work was focused on an original approach to investigate the hydrodynamics at different scales and then to intensify the performance of such complex reactors. The experiments were carried out respectively in a 3D reactor at macroscale, a 2D reactor at mesoscale and a 1D anaerobic reactor at microscale. A Particle Image Velocimetry (PIV), a micro-PIV and a high-speed camera were employed to quantify the liquid flow fields and the relative motion between sludge granules and bubbles. Shear rates exerted on sludge granules were quantified from liquid flow fields. The optimal biogas production is obtained at mean shear rate varying from 28 to 48s(-1), which is controlled by two antagonistic mechanisms. The multiscale approach demonstrates pertinent mechanisms proper to each scale and allows a better understanding of such reactors. Copyright © 2014 Elsevier Ltd. All rights reserved.

Status of the ENDF/B special applications files

International Nuclear Information System (INIS)

Stewart, L.

1977-01-01

The newly formed SAFE Subcommittee of the Cross Section Evaluation Working Group is charged with the responsibility for providing, reviewing, and testing several ENDF/B special purpose evaluated files. This responsibility currently encompasses dosimetry, activation, hydrogen and helium production, and radioactive decay data required by a variety of users. New formats have been approved by CSEWG for the inclusion of the activation and hydrogen and helium production cross-section libraries. The decay data will be in the same format as that already employed by the Fission Product and Actinide Subcommittee of CSEWG. While an extensive dosimetry file was available on the ENDF/B-IV library for fast reactor applications, other data are needed to extend the range of applications, especially to higher incident neutron energies. This Subcommittee has long-range plans to provide evaluated neutron interaction data that can be recommended for use in many specialized applications. 1 figure, 3 tables

Seismic analysis of fuel and target assemblies at a production reactor

International Nuclear Information System (INIS)

Braverman, J.I.; Wang, Y.K.

1991-01-01

This paper describes the unique modeling and analysis considerations used to assess the seismic adequacy of the fuel and target assemblies in a production reactor at Savannah River Site. This confirmatory analysis was necessary to provide assurance that the reactor can operate safely during a seismic event and be brought to a safe shutdown condition. The plant which was originally designed in the 1950's required to be assessed to more current seismic criteria. The design of the reactor internals and the magnitude of the structural responses enabled the use of a linear elastic dynamic analysis. A seismic analysis was performed using a finite element model consisting of the fuel and target assemblies, reactor tank, and a portion of the concrete structure supporting the reactor tank. The effects of submergence of the fuel and target assemblies in the water contained within the reactor tank can have a significant effect on their seismic response. Thus, the model included hydrodynamic fluid coupling effects between the assemblies and the reactor tank. Fluid coupling mass terms were based on formulations for solid bodies immersed in incompressible and frictionless fluids. The potential effects of gap conditions were also assessed in this evaluation. 5 refs., 6 figs., 1 tab

Facilities for the production and processing of radioisotopes

International Nuclear Information System (INIS)

Fourie, P.J.

1980-01-01

Radioisotopes which are used in South Africa are produced in the nuclear reactor SAFARI 1 of the AEB and the CSIR cyclotron in Pretoria or are being imported from various overseas manufactures. The safe and efficient production and use of radioisotopes is possible when being handled by sufficiently trained personnel using special designed equipment and facilities. The Isotope Production Centre is situated next to the reactor and waste treatment buildings. New production facilities shielded with lead and equipped with remote handling equipment are being erected and will be commissioned early during 1980 [af

JPRS Report, Science & Technology, Europe & Latin America, Argentina: Specifications of ARGOS 380 MW Reactor.

Science.gov (United States)

1988-02-18

and economy in operation. ENACE is an acronym for Empresa Nuclear Argentina de Centrales Electricas, or Argentine Nuclear Power Plant Corporation...products. The fuel transport system can be used during full reactor power operation to remove specially de - signed fuel assemblies containing rods for...SPECIFICATIONS OF ARGOS 380 MW REACTOR [Buenos Aires COMISION NACIONAL DE ENERGIA ATOMICA: INFORME in English 1987 pp 1-93] CONTENTS Argentine Offer of

Homogeneous SLOWPOKE reactor for the production of radio-isotope. A feasibility study

Energy Technology Data Exchange (ETDEWEB)

Busatta, P.; Bonin, H.W. [Royal Military College of Canada, Kingston, Ontario (Canada)]. E-mail: paul.busatta@rmc.ca; bonin-h@rmc.ca

2006-07-01

The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP Monte Carlo reactor calculation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether natural convection can still effectively cool the reactor using the modeling software FEMLAB. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)

Homogeneous SLOWPOKE reactor for the production of radio-isotope. A feasibility study

International Nuclear Information System (INIS)

Busatta, P.; Bonin, H.W.

2006-01-01

The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP Monte Carlo reactor calculation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether natural convection can still effectively cool the reactor using the modeling software FEMLAB. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)

Changes in water chemistry and primary productivity of a reactor cooling reservoir (Par Pond)

International Nuclear Information System (INIS)

Tilly, L.J.

1975-01-01

Water chemistry and primary productivity of a reactor cooling reservoir have been studied for 8 years. Initially the primary productivity increased sixfold, and the dissolved solids doubled. The dissolved-solids increase appears to have been caused by additions of makeup water from the Savannah River and by evaporative concentration during the cooling process. As the dissolved-solids concentrations and the conductivity of makeup water leveled off, the primary productivity stabilized. Major cation and anion concentrations generally followed total dissolved solids through the increase and plateau; however, silica concentrations declined steadily during the initial period of increased plankton productivity. Standing crops of net seston and centrifuge seston did not increase during this initial period. The collective data show the effects of thermal input to a cooling reservoir, illustrate the need for limnological studies before reactor siting, and suggest the possibility of using makeup-water additions to power reactor cooling basins as a reservoir management tool

Design of Continuous Reactor Systems for API Production

DEFF Research Database (Denmark)

Pedersen, Michael Jønch

-scale production equipment enabled complete replacement of the existing batch production of this intermediate. The crowning achievement in this work was the realization of continuous laboratory reactor setups capable of manufacturing the entire GMP portion of the synthesis of melitracen HCl at H. Lundbeck A....... The methodology provides suggestions on when and how decisions should be made on continuous production methods for Grignard chemistry within pharmaceutical manufacturing. Physicochemical properties, such as solubility, were found to be critical. However, from a business perspective, issues such as the current......The pharmaceutical industry has experienced many changes over the last few decades. Continuous production has been promoted as one of the more promising methods for making the industry more efficient and sustainable. The primary focus of this thesis is on the performance of Grignard chemistry...

New Production Reactors Program: Report to the Congress by the Secretary of Energy

International Nuclear Information System (INIS)

1990-01-01

In August 1988, DOE issued a report to Congress announcing the Department's proposal for assuring future tritium production capacity through the building of new facilities, specifically, a preferred duality strategy that calls for the design and construction of a new, up-to-date heavy water reactor (HWR) at the Savannah River Site in South Carolina; and, at the same time, to proceed with the design and construction of a modular high-temperature gas-cooled reactor (MHTGR) plant to be located at the Idaho National Engineering Laboratory in Idaho. As a contingency, the Department is developing a light water reactor (LWR) tritium target for possible use in the partially completed WNP-1 reactor located on the Hanford Site, near Richland, Washington. The Secretary created the Office of New Production Reactors (NP) to accomplish the following mission: To provide new production reactor capacity on an urgent schedule for an assured supply of nuclear materials, primarily tritium, to maintain the Nation's nuclear deterrent capability. The organizational philosophy of the Program calls for strong central control with increased safety, environmental, and quality assurance oversight. In addition to separate Engineering Technology Offices for HWR, MHTGR, and LWR that will maintain technical control throughout the life of the Program, the NP Director has created oversight Offices for Safety and Quality and for Environment. An NP Project Management Office will be established at each of the three project sites with the Project Officer reporting directly to the Director of NP. Safety, Quality, and Environmental Field Representatives will be located on-site at the Project Management Offices to serve as a ''resident'' for the Headquarters' Office to whom they will report

Nuclear reactor and production systems with digital controls

International Nuclear Information System (INIS)

Luger, P.P.

1976-01-01

Several digital sensing devices are described for use in automated production systems. The first described is for use in the automatic operation of a reactor. This device employs a binant electrometer using a quartz fiber mounted at one end but free to vibrate at the other in an AC field. The fiber oscillates if a charge is placed upon it. An optical slit replaces the ordinary eyepiece reticule scale. With the quartz fiber adjusted so its image is in focus at the optical slit, photoelectric signals are obtained at null charge on the fiber. The quartz fiber is repeatedly charged and allowed to discharge by collecting ions from a source under measurement. Each photoelectric signal causes a digital time reading to be taken. The time readings are used to evaluate the current due to the collected charge. The photoelectric signals, by feedback, also operate the electrometer for continuous or intermittent-continuous operation. Basically, the system is a current digitizer. Application is made to reactor monitoring and control as well as to other types of production systems. Finally, other types of sensing devices are also described and their use in automated controlled processes is shown. 3 claims, 19 figures

Cerenkov Detectors for Fission Product Monitoring in Reactor Coolant Water

Energy Technology Data Exchange (ETDEWEB)

Strindehag, O

1967-09-15

The expected properties of Cerenkov detectors when used for fission product monitoring in water cooled reactors and test loops are discussed from the point of view of the knowledge of the sensitivity of these detectors to some beta emitting isotopes. The basic theory for calculation of the detector response is presented, taking the optical transmission in the sample container and the properties of the photomultiplier tube into account. Special attention is paid to the energy resolution of this type of Cerenkov detector. For the design of practical detectors the results from several investigations of various window and reflector materials are given, and the selection of photomultiplier tubes is briefly discussed. In the case of optical reflectors and photomultiplier tubes reference is made to two previous reports by the author. The influence of the size and geometry of the sample container on the energy resolution follows from a separate investigation, as well as the relative merits of sample containers with transparent inner walls. Provided that the energy resolution of the Cerenkov detector is sufficiently high, there are several reasons for using this detector type for failed-fuel-element detection. It seems possible to attain the desired energy resolution by careful detector design.

Special irradiation techniques

International Nuclear Information System (INIS)

Colomez, Gerard; Veyrat, J.F.

1981-01-01

Irradiation trials conducted on materials-testing reactors should provide a better understanding of the phenomena which characterize the working and evolution in time of electricity-generating nuclear reactors. The authors begin by outlining the objectives behind experimental irradiation (applied to the various nuclear chains) and then describe the special techniques deployed to achieve these objectives [fr

Enhanced Passive Cooling for Waterless-Power Production Technologies

Energy Technology Data Exchange (ETDEWEB)

Rodriguez, Salvador B. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-06-14

Recent advances in the literature and at SNL indicate the strong potential for passive, specialized surfaces to significantly enhance power production output. Our exploratory computational and experimental research indicates that fractal and swirl surfaces can help enable waterless-power production by increasing the amount of heat transfer and turbulence, when compared with conventional surfaces. Small modular reactors, advanced reactors, and non-nuclear plants (e.g., solar and coal) are ideally suited for sCO2 coolant loops. The sCO2 loop converts the thermal heat into electricity, while the specialized surfaces passively and securely reject the waste process heat in an environmentally benign manner. The resultant, integrated energy systems are highly suitable for small grids, rural areas, and arid regions.

Production of polygalacturonases by Aspergillus oryzae in stirred tank and internal- and external-loop airlift reactors.

Science.gov (United States)

Fontana, Roselei Claudete; da Silveira, Maurício Moura

2012-11-01

The production of endo- and exo-polygalacturonase (PG) by Aspergillus oryzae was assessed in stirred tank reactors (STRs), internal-loop airlift reactors (ILARs) and external-loop airlift reactors (ELARs). For STR production, we compared culture media formulated with either pectin (WBE) or partially hydrolyzed pectin. The highest enzyme activities were obtained in medium that contained 50% pectin in hydrolyzed form (WBE5). PG production in the three reactor types was compared for WBE5 and low salt WBE medium, with additional salts added at 48, 60 and 72h (WBES). The ELARs performed better than the ILARs in WBES medium where the exo-PG was the same concentration as for STRs and the endo-PG was 20% lower. These results indicate that PG production is higher under experimental conditions that result in higher cell growth with minimum pH values less than 3.0. Copyright © 2012 Elsevier Ltd. All rights reserved.

The Y-12 National Security Complex Foreign Research Reactor Uranium Supply Production

Energy Technology Data Exchange (ETDEWEB)

Nelson, T. [Nuclear Technology and Nonproliferation Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States); Keller, A.P. [Disposition and Supply Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States)

2011-07-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the National Nuclear Security Administration (NNSA) HEU Disposition, the Reduced Enrichment Research and Test Reactors (RERTR), and the United States (U.S.) FRR Spent Nuclear Fuel (SNF) Acceptance Programs. The FRR Supply Program supports the important U.S. government nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to Low-Enriched Uranium (LEU) fuel under the RERTR Program. The NNSA Y-12 Site Office maintains the prime contracts with foreign government agencies for the supply of LEU for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. In addition to uranium metal feedstock for fuel fabrication, Y-12 can produce LEU in different forms to support new fuel development or target fabrication for medical isotope production. With production improvements and efficient delivery preparations, Y-12 continues to successfully support the global research reactor community. (author)

Production of 48V in a nuclear reactor via secondary tritons

International Nuclear Information System (INIS)

Siri, S.; Cohen, I.M.

2009-01-01

The production of 48 V in a nuclear reactor, induced on titanium by tritons generated from the 6 Li(n, t) 4 He reaction, and eventually 7 Li(n, n't) 4 He, is described. Samples of lithium titanate were irradiated for an irradiation cycle (120 h) in the RA-3 reactor, belonging to Ezeiza Atomic Centre. After a radiochemical separation, the characteristic radiations from 48 V were identified in the gamma ray spectra of the vanadium fractions. (orig.)

Optimization of hydrodynamic cavitations reactor efficiency for biodiesel production by response surface methods (Case study: Sunflower oil

Directory of Open Access Journals (Sweden)

H Javadikia

2017-05-01

Full Text Available Introduction Biofuels are considered as one of the largest sources of renewable fuels or replacement of fossil fuels. Combustion of plant-based fuels is the indirect use of solar energy. Biofuels significantly have less pollution than other fossil fuels and can easily generate from residual plant material. Waste and residues of foods and wastewater can also be a good source for biofuel production. Transesterification method (one of biodiesel production methods is the most common forms to produce mono-alkyl esters from vegetable oil and animal fats. The procedure aims are reduction the oil viscosity during the reaction between triglycerides and alcohol in the presence of a catalyst or without it. In this study, the method of transesterification with alkaline catalysts is used that it is the most common and most commercial biodiesel production method. In this study, configurations of made hydrodynamic cavitation reactor were studied to measure biodiesel fuel quality and enhanced device performance with optimum condition. The Design Expert software and response surface methodology were used to get this purpose. Materials and Methods Transesterification method was used in this study. The procedure aims were reduction of the oil viscosity during the reaction between triglycerides and alcohol in the presence of a catalyst or without it. Materials needed in the production of biodiesel transesterification method include: vegetable oil, alcohol and catalysts. The used oil in the production of biodiesel was sunflower oil, which was used 0.6 liters per each test in the production process base on titration method. Methanol with purity of 99.8 percent and the molar ratio of 6:1 to oil was used based on titration equation and according to the results of other researchers. The used catalyst in continuous production process was high-purity sodium hydroxide (99% that it is one of alkaline catalysts. Weight of hydroxide was 1% of the used oil weight in the

A New In-core Production Method of Co-60 in CANDU Reactors

Energy Technology Data Exchange (ETDEWEB)

Lyu, Jinqi; Kim, Woosong; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Park, Younwon [BEES Inc, Daejeon (Korea, Republic of)

2016-05-15

This study introduces an innovative method for Co-60 production in the CANDU6 core. In this new scheme, the central fuel element is replaced by a Co-59 target and Co-60 is obtained after the fuel bundle is discharged. It has been shown that the new method can produce significantly higher amount of Co-60 than the conventional Co production method in CANDU6 reactors without compromising the fuel burnup by removing some (<50%) of the adjuster rods in the whole core. The coolant void reactivity is noticeably reduced when a Co-59 target is loaded into the central pin of the fuel bundle. Meanwhile, the peak power in a fuel bundle is just a little higher due to the central Co-59 target than in conventional CANDU6 fuel design. The basic technology for Co-60 producing was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) in 1946 and the same technology was adapted and applied in CANDU6 power reactors. The standard CANDU6 reactor has 21 adjuster rods which are fully inserted into the core during normal operation. The stainless steel adjuster rods are replaced with neutronically-equivalent Co-59 adjusters to produce Co-60. Nowadays, the roles of the adjuster rods are rather vague since nuclear reactors cannot be quickly restarted after a sudden reactor trip due to more stringent regulations. In some Canadian CANDU6 reactors, some or all the adjuster rods are removed from the core to maximize the uranium utilization.

Reactor utilization, Annex A

International Nuclear Information System (INIS)

Martinc, R.; Stanic, A.

1984-01-01

Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

Recent advances on membranes and membrane reactors for hydrogen production

NARCIS (Netherlands)

Gallucci, F.; Fernandez Gesalaga, E.; Corengia, P.; Sint Annaland, van M.

2013-01-01

Membranes and membrane reactors for pure hydrogen production are widely investigated not only because of the important application areas of hydrogen, but especially because mechanically and chemically stable membranes with high perm-selectivity towards hydrogen are available and are continuously

Transmutation of long-lived fission product (137Cs, 90Sr) by a reactor-accelerator system

International Nuclear Information System (INIS)

Toyama, Shin-ichi; Takashita, Hirofumi; Konashi, Kenji; Sasao, Nobuyuki; Sato, Isamu.

1990-01-01

The report discusses the transmutation of long-lived fission products by a reactor and accelerator. It is important to take some criteria into consideration in transmutation disposal. To satisfy the criteria, a combined system of a reactor and an accelerator is proposed for the transmutation. An outline of the transmutation reactor and the accelerator is presented. The transmutation reactor has the ability to transmute a large quantity of fission products. However, it is desirable to have a high transmutation rate as well as a large disposal ability. Besides the transmutation property, it is necessary to investigate the physics of the transmutation reactor such as nuclear characteristics and burnup properties in order to obtain the most suitable, high performance core concept. A study on those properties is also presented. A high power accelerator is required for the transmutation. So a test linac is developed to accelerate high intensity beams. (N.K.)

Development and manufacturing of special fission chambers for in-core measurement requirements in nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Geslot, B.; Berhouet, F.; Oriol, L.; Breaud, S.; Jammes, C.; Filliatre, P.; Villard, J. F. [CEA, DEN, Dosimetry Command Control and Instrumentation Laboratory, F-13109 Saint-Paul-lez-Durance (France)

2009-07-01

The Dosimetry Command control and Instrumentation Laboratory (LDCI) at CEA/Cadarache is specialized in the development, design and manufacturing of miniature fission chambers (from 8 mm down to 1.5 mm in diameter). The LDCI fission chambers workshop specificity is its capacity to manufacture and distribute special fission chambers with fissile deposits other than U{sup 235} (typically Pu{sup 242}, Np{sup 237}, U{sup 238}, Th{sup 232}). We are also able to define the characteristics of the detector for any in-core measurement requirements: sensor geometry, fissile deposit material and mass, filling gas composition and pressure, operating mode (pulse, current or Campbelling) with associated cable and electronics. The fission chamber design relies on numerical simulation and modeling tools developed by the LDCI. One of our present activities in fission chamber applications is to develop a fast neutron flux instrumentation using Campbelling mode dedicated to measurements in material testing reactors. (authors)

Simulation of corrosion product activity in pressurized water reactors under flow rate transients

International Nuclear Information System (INIS)

Mirza, Anwar M.; Mirza, Nasir M.; Mir, Imran

1998-01-01

Simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been done under flow rate transients. Employing time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for 24 Na, 56 Mn, 59 Fe, 60 Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to Mn-56. Parametric studies at full power for various ramp decreases in flow rate show initial decline in the activity and then a gradual rise to relatively higher saturation values. The minimum value and the time taken to reach the minima are strong functions of the slope of linear decrease in flow rate. In the second part flow rate coastdown was allowed to occur at different flow half-times. The reactor scram was initiated at 90% of the normal flow rate. The results show that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions

The University of Missouri Research Reactor, its fuel and productivity

International Nuclear Information System (INIS)

Brugger, R.M.; Schlapper, G.A.; Alger, D.M.

1993-01-01

This paper describes the University of Missouri Research Reactor (MURR) and presents a summary of contributions to education, research, and service. These efforts have helped offset the impact on the U.S. economy of research from other countries. Special emphasis is placed on fuel design developments that have allowed MuRR to keep the cost megawatt day (MWD) of fuel essentially constant. Also noted is the fact that the United States has missed some research opportunities because of a hold-the-line attitude. The slipping position of U.S. research reactors is compared with the rest of the world. As will be further outlined in the text, the MURR cannot (with available technology) decrease its U-235 enrichment level and maintain present research capabilities and fuel cycle costs. Data is presented to show how MURR, if permitted to use advanced fuel technology, could reduce fuel fabrication costs and onsite U-235 inventory. In addition it is shown that MURR could increase its capabilities provided that arbitrary institutional limits are removed

The University of Missouri Research Reactor, its fuel and productivity

International Nuclear Information System (INIS)

Brugger, Robert M.; Schlapper, Gerald A.; Alger, Don M.

1993-01-01

This paper describes the University of Missouri Research Reactor (MURR) and presents a summary of contributions to education, research, and service. These efforts have helped offset the impact on the U. S. economy of research from other countries. Special emphasis is placed on fuel design developments that have allowed MURR to keep the cost megawatt day (MWD) of fuel essentially constant. Also noted is the fact that the United States has missed some research opportunities because of a hold-the-line attitude. The slipping position of U.S. research reactors is compared with the rest of the world. As will be further outlined in the text, the MURR cannot (with available technology) decrease its U-235 enrichment level and maintain present research capabilities and fuel cycle costs. Data is presented to show how MURR if permitted to use advanced fuel technology, could reduce fuel fabrication costs and onsite U-235 inventory. In addition it is shown that MURR could increase its capabilities provided that arbitrary institutional limits are removed

Power unit with GT-MHR reactor plant for electricity production and district heating

International Nuclear Information System (INIS)

Kiryushin, A.L.; Kodochigov, N.G.; Kuzavkov, N.G.; Golovko, V.F.

2000-01-01

Modular helium reactor with the gas turbine (GT-MHR) is a perspective power reactor plant for the next century. The project reactor is based on experience of operation more than 50 gas-cooled reactors on carbon dioxide and helium, and also on subsequent achievements in the field of realization direct gas turbine Brayton cycle. To the beginning of 90 years, achievements in technology of gas turbines, highly effective recuperators and magnetic bearings made it possible to start development of the reactor plant project combining a safe modular gas cooled reactor and a power conversion system, realizing the highly effective Brayton cycle. The conceptual project of the commercial GT-MHR reactor plant fulfilled in 1997 by joint efforts of international firms, combines a safe modular reactor with an annular active core of prismatic fuel blocks and a power conversion system with direct gas turbine cycle. The efficiency of GT-MHR gas turbine cycle at level of about 48% makes it competitive in the electricity production market in comparison with any fossil or nuclear power stations

Draft environmental impact statement siting, construction, and operation of New Production Reactor capacity. Volume 4, Appendices D-R

Energy Technology Data Exchange (ETDEWEB)

None

1991-04-01

This Environmental Impact Statement (EIS) assesses the potential environmental impacts, both on a broad programmatic level and on a project-specific level, concerning a proposed action to provide new tritium production capacity to meet the nation`s nuclear defense requirements well into the 21st century. A capacity equivalent to that of about a 3,000-megawatt (thermal) heavy-water reactor was assumed as a reference basis for analysis in this EIS; this is the approximate capacity of the existing production reactors at DOE`s Savannah River Site near Aiken, South Carolina. The EIS programmatic alternatives address Departmental decisions to be made on whether to build new production facilities, whether to build one or more complexes, what size production capacity to provide, and when to provide this capacity. Project-specific impacts for siting, constructing, and operating new production reactor capacity are assessed for three alternative sites: the Hanford Site near Richland, Washington; the Idaho National Engineering Laboratory near Idaho Falls, Idaho; and the Savannah River Site. For each site, the impacts of three reactor technologies (and supporting facilities) are assessed: a heavy-water reactor, a light-water reactor, and a modular high-temperature gas-cooled reactor. Impacts of the no-action alternative also are assessed. The EIS evaluates impacts related to air quality; noise levels; surface water, groundwater, and wetlands; land use; recreation; visual environment; biotic resources; historical, archaeological, and cultural resources; socioeconomics; transportation; waste management; and human health and safety. The EIS describes in detail the potential radioactive releases from new production reactors and support facilities and assesses the potential doses to workers and the general public. This volume contains 15 appendices.

Numerical simulation of vortex pyrolysis reactors for condensable tar production from biomass

Energy Technology Data Exchange (ETDEWEB)

Miller, R.S.; Bellan, J. [California Inst. of Tech., Pasadena, CA (United States). Jet Propulsion Lab.

1998-08-01

A numerical study is performed in order to evaluate the performance and optimal operating conditions of vortex pyrolysis reactors used for condensable tar production from biomass. A detailed mathematical model of porous biomass particle pyrolysis is coupled with a compressible Reynolds stress transport model for the turbulent reactor swirling flow. An initial evaluation of particle dimensionality effects is made through comparisons of single- (1D) and multi-dimensional particle simulations and reveals that the 1D particle model results in conservative estimates for total pyrolysis conversion times and tar collection. The observed deviations are due predominantly to geometry effects while directional effects from thermal conductivity and permeability variations are relatively small. Rapid ablative particle heating rates are attributed to a mechanical fragmentation of the biomass particles that is modeled using a critical porosity for matrix breakup. Optimal thermal conditions for tar production are observed for 900 K. Effects of biomass identity, particle size distribution, and reactor geometry and scale are discussed.

Flexible robotic entry device for a nuclear materials production reactor

International Nuclear Information System (INIS)

Heckendorn, F.M. II.

1988-01-01

The Savannah River Laboratory has developed and is implementing a flexible robotic entry device (FRED) for the nuclear materials production reactors now operating at the Savannah River Plant (SRP). FRED is designed for rapid deployment into confinement areas of operating reactors to assess unknown conditions. A unique smart tether method has been incorporated into FRED for simultaneous bidirectional transmission of multiple video/audio/control/power signals over a single coaxial cable. This system makes it possible to use FRED under all operating and standby conditions, including those where radio/microwave transmissions are not possible or permitted, and increases the quantity of data available

A gas-phase reactor powered by solar energy and ethanol for H2 production

International Nuclear Information System (INIS)

Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

2014-01-01

In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

International Nuclear Information System (INIS)

Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez; Universidade Federal de Pernambuco

2017-01-01

The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly "9"9Mo. Compare to multipurpose research reactors, an AHR dedicated for "9"9Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

Energy Technology Data Exchange (ETDEWEB)

Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez, E-mail: milianperez89@gmail.com, E-mail: dmilian@instec.cu, E-mail: lorenapilar1109@gmail.com, E-mail: cabol@ufpe.br [Higher Institute of Technologies and Applied Sciences (InSTEC), Havana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

2017-11-01

The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly {sup 99}Mo. Compare to multipurpose research reactors, an AHR dedicated for {sup 99}Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

Commercial products and services of research reactors. Proceedings of a technical meeting

International Nuclear Information System (INIS)

2013-07-01

Although the number of operational research reactors is steadily decreasing, more than half of those that remain are greatly underutilized and, in most cases, underfunded. To continue to play a key role in the development of peaceful uses of nuclear technology, the remaining research reactors will need to provide useful products and services to private, national and regional customers, in some cases with adequate revenue generation for reliable, safe and secure facility management and operation. In the light of declining governmental financial support and the need for improved physical security and conversion to low enriched uranium (LEU) fuel, many research reactors have been challenged to generate income to offset increasing operational and maintenance costs. The renewed interest in nuclear power (and therefore in nuclear education and training), the global expansion of diagnostic and therapeutic nuclear medicine, and the extensive use of semiconductors in electronics and in other areas have created new opportunities for research reactors, prominent among them, markets for products and services in regions and countries without such facilities. It is clear that such initiatives towards greater self-reliance will need to address such aspects as market surveys, marketing and business plans, and cost of delivery services. It will also be important to better inform present and future potential end users of research reactor services of the capabilities and products that can be provided. This publication is a compilation of material from an IAEA technical meeting on “Commercial Products and Services of Research Reactors”, held in Vienna, Austria, from 28 June to 2 July 2010. The overall objective of the meeting was to exchange information on good practices and to provide concrete examples, in technical presentations and brainstorming discussions, to promote and facilitate the development of commercial applications of research reactors. The meeting also aimed to

Advanced reactor development for non-electric applications

International Nuclear Information System (INIS)

Chang, M.H.; Kim, S.H.

1996-01-01

Advance in the nuclear reactor technology achieved through nuclear power programs carried out in the world has led nuclear communities to direct its attention to a better and peaceful utilization of nuclear energy in addition to that for power generation. The efforts for non-electric application of nuclear energy has been pursued in a limited number of countries in the world for their special needs. However, those needs and the associated efforts contributed largely to the development and practical realization of advanced reactors characterized by highly improved reactor safety and reliability by deploying the most up-to-date safety technologies. Due mainly to the special purpose of utilization, economic reasons and ease in implementation of new advanced technologies, small and medium reactors have become a major stream in the reactor developments for non-electric applications. The purpose of this paper is to provide, to the interested nuclear society, the overview of the development status and design characteristics of selected advanced nuclear reactors previously developed and/or currently under development specially for non-electric applications. Major design technologies employed in those reactors to enhance the reactor safety and reliability are reviewed to present the underlying principles of the design. Along with the overview, this paper also introduces a development program and major design characteristics of an advanced integral reactor (SMART) for co-generation purpose currently under conceptual development in Korea. (author)

The reactor and the production of isotopes

International Nuclear Information System (INIS)

Hevesy, G. de

1962-01-01

The construction of the cyclotron immensely advanced the availability of radioactive tracers, a few of which even today can be produced only with the aid of this device. But even this great advance was overshadowed by the fabulous production of isotopes by the reactors. Isotopes of almost any element and of almost unlimited activity became available. It now became possible to apply H 3 - discovered already in the 'thirties by Rutherford and Oliphant - and C 14 , and these were used in thousands of investigations

International seminar on status and prospects for small and medium sized reactors. Book of extended synopses

International Nuclear Information System (INIS)

2001-05-01

Small and medium sized reactors have been defined by IAEA in accordance with their net electrical power rating, or thermal equivalent power rating for reactors producing process heat. Within the global energy supply and demand the role of nuclear power, and specifically the role of small and medium sizes reactors, is reviewed. Some new design prospects of dual purpose power plants, namely for heat production and desalination, are presented. Some proposal for new concepts of small size power reactors are described. A special concern is devoted to economic and energy policy issues as well as to safety and non-proliferation matters

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

Science.gov (United States)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

Processes of hydrogen production, coupled with nuclear reactors: Economic perspectives

International Nuclear Information System (INIS)

Werkoff, Francois; Avril, Sophie; Mansilla, Christine; Sigurvinsson, Jon

2006-01-01

Hydrogen production, using nuclear power is considered from a technic-economic (TE) point of view. Three different processes are examined: Alkaline electrolysis, High-temperature steam electrolysis (HTE) and the thermochemical Sulphur-Iodine (S/I) cycle. The three processes differ, in the sense that the first one is operational and both last ones are still at demonstration stages. For them, it is at present only possible to identify key points and limits of competitiveness. The cost of producing hydrogen by alkaline electrolysis is analysed. Three major contributions to the production costs are examined: the electricity consumption, the operation and maintenance expenditures and the depreciation capital expenditures. A technic-economic evaluation of hydrogen production by HTE coupled to a high-temperature reactor (HTR) is presented. Key points appear to be the electrolyser and the high temperature heat exchangers. The S/I thermochemical cycle is based on the decomposition and the re-composition of H 2 SO 4 and HI acids. The energy consumption and the recovery of iodine are key points of the S/I cycle. With the hypothesis that the hydrogen energy will progressively replace the fossil fuels, we give a first estimate of the numbers of nuclear reactors (EPR or HTR) that would be needed for a massive nuclear hydrogen production. (authors)

Use of dwell time concept in fission product inventory assessment for CANDU reactors

International Nuclear Information System (INIS)

Bae, C.J.; Choi, J.H.; Hwang, H.R.; Seo, J.T.

2003-01-01

A realistic approach in calculating the initial fission product inventory within the CANFLEX-NU fuel has been assessed for its applicability to the single channel event safety analysis for CANDU reactors. This approach is based on the dwell time concept in which the accident is assumed to occur at the dwell time when the summation of fission product inventory for all isotopes becomes largest. However, in the current conservative analysis, the maximum total inventory and the corresponding gap inventory for each isotope are used as the initial fission product inventories regardless of the accident initiation time. The fission product inventory analysis has been performed using ELESTRES code considering power histories and burnup of the fuel bundles in the limiting channel. The analysis results showed that the total fission product inventory is found to be largest at 20% dwell time. Therefore, the fission product inventory at 20% dwell time can be used as the initial condition for the single channel event for the CANDU 6 reactors. (author)

Analysis of 99Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor using ORIGEN and MCNP

Directory of Open Access Journals (Sweden)

A. Isnaeni

2014-04-01

Full Text Available 99mTc is a very useful radioisotope in medical diagnostic procedure. 99mTc is produced from 99Mo decay. Currently, most of 99Mo is produced by irradiating 235U in the nuclear reactor. 99Mo mostly results from the fission reaction of 235U targets with a fission yield about 6.1%. A small additional amount is created from 98Mo neutron activation. Actually 99Mo is also created in the reactor fuel, but usually we do not extract it. The fuel will become spent fuel which is a highly radioactive waste. 99Mo production system in the aqueous homogeneous reactor offers a better method, because all of the 99Mo can be extracted from the fuel solution. Fresh reactor fuel solution consists of uranyl nitrate dissolved in water. There is no separation of target and fuel in an aqueous homogeneous reactor where target and fuel become one liquid solution, and there is no spent fuel generated from this reactor. Simulation of the extraction process is performed while reactor in operation (without reactor shutdown. With an extraction flow rate of 3.6 L/h, after 43 hours of reactor operation the production of 99Mo is relatively constant at about 98.6 curie/hour

48 CFR 1545.309 - Providing Government production and research property under special restrictions.

Science.gov (United States)

2010-10-01

... improvements necessary for installing special tooling, special test equipment, or plant equipment, shall not be... production and research property under special restrictions. 1545.309 Section 1545.309 Federal Acquisition Regulations System ENVIRONMENTAL PROTECTION AGENCY CONTRACT MANAGEMENT GOVERNMENT PROPERTY Providing...

Removal of decay heat by specially designed isolation condensers for advanced heavy water reactors

Energy Technology Data Exchange (ETDEWEB)

Dhawan, M L; Bhatia, S K [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

For Advanced Heavy Water Reactor (AHWR), removal of decay heat and containment heat is being considered by passive means. For this, special type of isolation condensers are designed. Isolation condensers when submerged in a pool of water, are the best choice because condensation of high temperature steam is an extremely efficient heat transfer mechanism. By the use of isolation condensers, not only heat is removed but also pressure and temperature of the system are automatically controlled without losing the coolant and without using conventional safety relief valves. In this paper, design optimisation studies of isolation condensers of different types with natural circulation for the removal of core decay heat for AHWR is presented. (author). 8 refs., 2 figs.

Production of Sn-117m in the BR2 high-flux reactor.

Science.gov (United States)

Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

2009-01-01

The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

Reactor production and processing of radioisotopes for therapeutic applications in nuclear medicine

International Nuclear Information System (INIS)

Knapp, F.F. Jr.; Mirzadeh, S.; Beets, A.L.

1995-01-01

Nuclear reactors continue to play an important role in providing radioisotopes for nuclear medicine. Many reactor-produced radioisotopes are ''neutron rich'' and decay by beta-emission and are thus of interest for therapeutic applications. This talk discusses the production and processing of a variety of reactor-produced radioisotopes of current interest, including those produced by the single neutron capture process, double neutron capture and those available from beta-decay of reactorproduced radioisotopes. Generators prepared from reactorproduced radioisotopes are of particular interest since repeated elution inexpensively provides many patient doses. The development of the alumina-based W-188/Re-188 generator system is discussed in detail

PSA in design of passive/active safety reactors

International Nuclear Information System (INIS)

Sato, T.; Tanabe, A.; Kondo, S.

1995-01-01

PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

Fast pyrolysis of Miscanthus sinensis in fluidized bed reactors: Characteristics of product yields and biocrude oil quality

International Nuclear Information System (INIS)

Bok, Jin Pil; Choi, Hang Seok; Choi, Joon Weon; Choi, Yeon Seok

2013-01-01

In the present work, fast pyrolysis of Miscanthus sinensis was performed and the product yields and properties of the resulting biocrude oil were determined for varying reactor configurations and pyrolysis temperatures. Two types of reactors (rectangular and cylindrical fluidized beds) were adopted, and pyrolysis temperature was increased from 400 °C to 550 °C. Based on the results, it was found that the reaction temperature greatly influenced the product yield and the characteristics of biocrude oil. The highest yield of biocrude oil for the rectangular reactor was 48.9 wt.%, produced at 500 °C, and the highest yield for the cylindrical reactor was 50.01 wt.%, produced at 450 °C. Additionally, the biocrude oil yield in the rectangular reactor sharply decreased when reaction temperature was increased to 550 °C, while only a slight decrease was observed in the cylindrical reactor. From GC/MS analysis, biocrude oil was found to contain various chemical components, such as nonaromatic ketones, furans, sugars, lignin-derived phenols, guaiacols and syringols. In particular, the sugar content of the biocrude oil produced in rectangular reactor (2.11–9.35 wt.%) was generally lower than that produced in the cylindrical reactor (7.93–10.79 wt.%). - Highlights: • Fast pyrolysis of Miscanthus sinensis was performed in two fluidized bed reactors to obtain biocrude oil. • The yield and characteristics of the biocrude oil were scrutinized with changing reaction temperature and reactor type. • The reaction temperature was found to be the most influencing parameter for the fast pyrolysis reaction. • The different heating rate caused by reactor type has an effect on the final product yield and characteristics

Directions of Development of International Production Specialization Theory in Terms of Integration

Directory of Open Access Journals (Sweden)

Lidiya Aleksandrovna Pankova

2016-06-01

Full Text Available The article is devoted to the definition of directions of development of the theory of international production specialization in terms of integration based on new trends in various areas of international business. The author proposes a new concept of international industrial specialization in terms of integration. As a result of research the author came to the conclusion that in the current economic conditions the need for a free market constraints associated with the deepening global market failures is becoming more and more obvious. Therefore, in the process of international specialization of production in the conditions of integration of the country should be guided not by comparative advantage in the production of certain goods or services, but also by other criteria. These criteria are: financial criterion determining the financial situation in the country, the trade criterion, reflecting the possibility of benefiting from international trade, technological criterion, considering the technological security of the country in the field of specialization and the cyclical criterion, reflecting the situation in the world economy and the level of integration of the country. On the basis of these criteria, the author identifies three categories of countries according to their ability to benefit from an international industrial specialization in the integration process: effective export, export-neutral and export-inefficient countries.

Manufacture of components for Canadian reactor programs

International Nuclear Information System (INIS)

Perry, L.P.

Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 8, Pomocni sistemi

Energy Technology Data Exchange (ETDEWEB)

NONE

1986-11-01

This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport. [Serbo-Croat] Ova knjiga obuhvata opis pomocnih sistem reaktora RA: sistem specijalne ventilacije, sistem specijalne kanalizacije, vruce komore, sistemi za unutrasnji transport. Ventilacioni sistem je znacajan deo sistema zastite i sigurnosti reaktora. Njegova je uloga da onemoguci disperziju radioaktivnih cestica u okolinu. Sistem specijalne kanalizacije sastoji se od agregata, cevovoda i rezervoara sa sigurnosnom armaturom i ima zadatak da prihvata i u sebe deponuje radioaktivne otpadne vode iz objekarta reaktora RA. U zgradi reaktora RA postoje vruce komore koje su namenjene za proizvodnju zatvorenih izvora zracenje, ukljucujuci i mehanicku obradu, prepakivanje i transport.

Energy production from distillery wastewater using single and double-phase upflow anaerobic sludge blanket (UASB) reactor

Energy Technology Data Exchange (ETDEWEB)

Muyodi, F J; Rubindamayugi, M S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

1998-12-31

A Single-phase (SP) and Double-phase (DP) Upflow Anaerobic Sludge Blanket (UASB) reactors treating distillery wastewater were operated in parallel. The DP UASB reactor showed better performance than the SP UASB reactor in terms of maximum methane production rate, methane content and Chemical Oxygen Demand (COD) removal efficiency. (au) 20 refs.

Energy production from distillery wastewater using single and double-phase upflow anaerobic sludge blanket (UASB) reactor

Energy Technology Data Exchange (ETDEWEB)

Muyodi, F.J.; Rubindamayugi, M.S.T. [Univ. of Dar es Salaam, Applied Microbiology Unit (Tanzania, United Republic of)

1997-12-31

A Single-phase (SP) and Double-phase (DP) Upflow Anaerobic Sludge Blanket (UASB) reactors treating distillery wastewater were operated in parallel. The DP UASB reactor showed better performance than the SP UASB reactor in terms of maximum methane production rate, methane content and Chemical Oxygen Demand (COD) removal efficiency. (au) 20 refs.

Energy production and reactor efficiency

International Nuclear Information System (INIS)

Anon.

1975-01-01

Doubts have been raised in relation to the economic and energetic efficiency of nuclear reactors. Some economists are questioning whether, when all the capital and material inputs to fission technology are considered, nuclear reactors yield sufficiently large amounts of energy to show a nett gain of energy. (author)

RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1984-12-01

During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

International Nuclear Information System (INIS)

Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

1985-01-01

According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

Method and device for controlling nuclear reactor power

International Nuclear Information System (INIS)

Takigawa, Yukio; Ebata, Shigeo.

1988-01-01

Purpose: To detect and suppress the special power oscillations in the reactor core. Method: Four pairs of LPRM detectors, each pair comprising two detectors are disposed at an identical axial direction of the reactor core and situated at substantially insymmetrical positions at least in longitudinal, vertical and orthogonal directions with respect to the center of te reactor core and LPRM signals from them are inputted into a device for judging special power oscillations. In this case, a standardized mutual relation function is determined on every pair for the respective LPRM signals. Generation of special power oscillations in the reactor core is judged when it is detected that peaks appearing at least in one of the function forms for each pair are negative and have absolute values exceeding a predetermined value and that time of peak is within a predetermined time. The judged signal is inputted to a selected control rod insertion device. The selected control rod insertion device, upon preceiving the signal, inserts selected control rods into the reactor core to suppress the special power oscillations. Accordingly, it is possible to improve the fuel integrity. (Horiuchi, T.)

Design and construction of a 7,500 liter immobilized cell reactor-separator for ethanol production from whey

Energy Technology Data Exchange (ETDEWEB)

Dale, M.C.

1992-12-31

A 7,500 liter reactor/separator has been constructed for the production of ethanol from concentrated whey permeate. This unit is sited in Hopkinton IA, across the street from a whey generating cheese plant A two phase construction project consisting of (1) building and testing a reactor/separator with a solvent absorber in a single unified housing, and (2) building and testing an extractive distillation/product stripper for the recovery of anhydrous ethanol is under way. The design capacity of this unit is 250,000 gal/yr of anhydrous product. Design and construction details of the reactor/absorber separator are given, and design parameters for the extractive distillation system are described.

Novolak Production

Science.gov (United States)

Aiba, Hiroshi

Novolak resins are produced by reacting formaldehyde (30-55% concentration) with phenol under acidic conditions, with oxalic acid as the preferred catalyst and in special conditions, sulfuric acid. Depending on the batch size, all raw material components can be introduced into the reactor, or when there is an increase in the batch size as well as in the reactor volume, the reaction exotherm is controlled by a gradual addition of formaldehyde. Modern novolak production facilities are automated and programmed for reduced operational cost. A flow diagram of a general production line for the manufacture of novolak is shown. Recovery of the novolak is accomplished by the removal of water and devolatilization of crude novolak to molten, low-free phenol novolak resin which can be isolated as flake or pastille or dissolved in appropriate solvents. Novolak is stored either in a solid flake or pastille form or in solution. Most production is conducted under atmospheric conditions, but there are some recent, novel activities such as pressure in a hermetically-closed reactor reaching 0.1-10 MPa by using the heat of reaction without reflux to shorten reaction time, accelerating dehydration time by flash distillation, and providing economic benefit in the cost of novolak production.

Development status on hydrogen production technology using high-temperature gas-cooled reactor at JAEA, Japan

International Nuclear Information System (INIS)

Shiozawa, Shusaku; Ogawa, Masuro; Hino, Ryutaro

2006-01-01

The high-temperature gas-cooled reactor (HTGR), which is graphite-moderated and helium-cooled, is attractive due to its unique capability of producing high temperature helium gas and its fully inherent reactor safety. In particular, hydrogen production using the nuclear heat from HTGR (up to 900 deg. C) offers one of the most promising technological solutions to curb the rising level of CO 2 emission and resulting risk of climate change. The interests in HTGR as an advanced nuclear power source for the next generation reactor, therefore, continue to rise. This is represented by the Japanese HTTR (High-Temperature Engineering Test Reactor) Project and the Chinese HTR-10 Project, followed by the international Generation IV development program, US nuclear hydrogen initiative program, EU innovative HTR technology development program, etc. To enhance nuclear energy application to heat process industries, the Japan Atomic Energy Agency (JAEA) has continued extensive efforts for development of hydrogen production system using the nuclear heat from HTGR in the framework of the HTTR Project. The HTTR Project has the objectives of establishing both HTGR technology and heat utilization technology. Using the HTTR constructed at the Oarai Research and Development Center of JAEA, reactor performance and safety demonstration tests have been conducted as planned. The reactor outlet temperature of 950 deg. C was successfully achieved in April 2004. For hydrogen production as heat utilization technology, R and D on thermo-chemical water splitting by the 'Iodine-Sulfur process' (IS process) has been conducted step by step. Proof of the basic IS process was made in 1997 on a lab-scale of hydrogen production of 1 L/h. In 2004, one-week continuous operation of the IS process was successfully demonstrated using a bench-scale apparatus with hydrogen production rate of 31 L/h. Further test using a pilot scale facility with greater hydrogen production rate of 10 - 30 m 3 /h is planned as

Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

International Nuclear Information System (INIS)

Ipek, O.

1997-01-01

The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

Non-catalytic alcoholysis process for production of biodiesel fuel by using bubble column reactor

Science.gov (United States)

Hagiwara, S.; Nabetani, H.; Nakajima, M.

2015-04-01

Biodiesel fuel is a replacement for diesel as a fuel produced from biomass resources. It is usually defined as a fatty acid methyl ester (FAME) derived from vegetable oil or animal fat. In European countries, such as Germany and France, biodiesel fuel is commercially produced mainly from rapeseed oil, whereas in the United States and Argentina, soybean oil is more frequently used. In many other countries such as Japan and countries in Southeast Asia, lipids that cannot be used as a food source could be more suitable materials for the production of biodiesel fuel because its production from edible oils could result in an increase in the price of edible oils, thereby increasing the cost of some foodstuffs. Therefore, used edible oil, lipids contained in waste effluent from the oil milling process, byproducts from oil refining process and crude oils from industrial crops such as jatropha could be more promising materials in these countries. The materials available in Japan and Southeast Asia for the production of biodiesel fuel have common characteristics; they contain considerable amount of impurities and are high in free fatty acids (FFA). Superheated methanol vapor (SMV) reactor might be a promising method for biodiesel fuel production utilizing oil feedstock containing FFA such as waste vegetable oil and crude vegetable oil. In the conventional method using alkaline catalyst, FFA contained in waste vegetable oil is known to react with alkaline catalyst such as NaOH and KOH generating saponification products and to inactivate it. Therefore, the FFA needs to be removed from the feedstock prior to the reaction. Removal of the alkaline catalyst after the reaction is also required. In the case of the SMV reactor, the processes for removing FFA prior to the reaction and catalyst after the reaction can be omitted because it requires no catalyst. Nevertheless, detailed study on the productivity of biodiesel fuel produced from waste vegetable oils and other non

Non-catalytic alcoholysis process for production of biodiesel fuel by using bubble column reactor

International Nuclear Information System (INIS)

Hagiwara, S; Nabetani, H; Nakajima, M

2015-01-01

Biodiesel fuel is a replacement for diesel as a fuel produced from biomass resources. It is usually defined as a fatty acid methyl ester (FAME) derived from vegetable oil or animal fat. In European countries, such as Germany and France, biodiesel fuel is commercially produced mainly from rapeseed oil, whereas in the United States and Argentina, soybean oil is more frequently used. In many other countries such as Japan and countries in Southeast Asia, lipids that cannot be used as a food source could be more suitable materials for the production of biodiesel fuel because its production from edible oils could result in an increase in the price of edible oils, thereby increasing the cost of some foodstuffs. Therefore, used edible oil, lipids contained in waste effluent from the oil milling process, byproducts from oil refining process and crude oils from industrial crops such as jatropha could be more promising materials in these countries. The materials available in Japan and Southeast Asia for the production of biodiesel fuel have common characteristics; they contain considerable amount of impurities and are high in free fatty acids (FFA). Superheated methanol vapor (SMV) reactor might be a promising method for biodiesel fuel production utilizing oil feedstock containing FFA such as waste vegetable oil and crude vegetable oil. In the conventional method using alkaline catalyst, FFA contained in waste vegetable oil is known to react with alkaline catalyst such as NaOH and KOH generating saponification products and to inactivate it. Therefore, the FFA needs to be removed from the feedstock prior to the reaction. Removal of the alkaline catalyst after the reaction is also required. In the case of the SMV reactor, the processes for removing FFA prior to the reaction and catalyst after the reaction can be omitted because it requires no catalyst. Nevertheless, detailed study on the productivity of biodiesel fuel produced from waste vegetable oils and other non

Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

Energy Technology Data Exchange (ETDEWEB)

Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

2013-01-29

Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

Very High Efficiency Reactor (VHER) Concepts for Electrical Power Generation and Hydrogen Production

International Nuclear Information System (INIS)

PARMA JR, EDWARD J.; PICKARD, PAUL S.; SUO-ANTTILA, AHTI JORMA

2003-01-01

The goal of the Very High Efficiency Reactor study was to develop and analyze concepts for the next generation of nuclear power reactors. The next generation power reactor should be cost effective compared to current power generation plant, passively safe, and proliferation-resistant. High-temperature reactor systems allow higher electrical generating efficiencies and high-temperature process heat applications, such as thermo-chemical hydrogen production. The study focused on three concepts; one using molten salt coolant with a prismatic fuel-element geometry, the other two using high-pressure helium coolant with a prismatic fuel-element geometry and a fuel-pebble element design. Peak operating temperatures, passive-safety, decay heat removal, criticality, burnup, reactivity coefficients, and material issues were analyzed to determine the technical feasibility of each concept

Assessment of radiation fields from neutron irradiated structural components of the 40 MW research reactor CIRUS

International Nuclear Information System (INIS)

Sankaranarayanan, S.; Sharma, S.K.

1993-01-01

The paper summarizes the results of an assessment of the radiation fields from the long-lived neutron activation products (including the decay chain products) in the various structural components of the CIRUS reactor. Special attention is given for the analysis of neutron activation of impurity elements present in the materials of the structure. 16 refs, 4 figs, 4 tabs

Design of a multipurpose research reactor

International Nuclear Information System (INIS)

Sanchez Rios, A.A.

1990-01-01

The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute

Comparison of the production of medical radioisotopes on reactor and cyclotron

International Nuclear Information System (INIS)

Vucina, J.; Vuksanovic, Lj.; Dobrijevic, R.; Karanfilov, E.

1997-01-01

The production of radioisotopes for nuclear-medical applications can be performed either on nuclear reactor or on cyclotron. According to the nuclear reactions applied the radioisotopes of different physical characteristics can be produced. In the paper a comparison of the radioisotopes production given. Compared are the main steps in the production: choice of the nuclear reaction, targetry, irradiation and radiochemical separations performed on the irradiated target to isolate the desired radioisotope. The main characteristics of the produced radioisotopes are given and discussed. (author)

Biohydrogen production from pig slurry in a CSTR reactor system with mixed cultures under hyper-thermophilic temperature (70 oC)

International Nuclear Information System (INIS)

Kotsopoulos, Thomas A.; Fotidis, Ioannis A.; Tsolakis, Nikolaos; Martzopoulos, Gerassimos G.

2009-01-01

A continuous stirred tank reactor (CSTR) (750 cm 3 working volume) was operated with pig slurry under hyper-thermophilic (70 o C) temperature for hydrogen production. The hydraulic retention time (HRT) was 24 h and the organic loading rate was 24.9 g d -1 of volatile solid (VS). The inoculum used in the hyper-thermophilic reactor was sludge obtained from a mesophilic methanogenic reactor. The continuous feeding with active biomass (inoculum) from the mesophilic methanogenic reactor was necessary in order to achieve hydrogen production. The hyper-thermophilic reactor started to produce hydrogen after a short adapted period of 4 days. During the steady state period the mean hydrogen yield was 3.65 cm 3 g -1 of volatile solid added. The high operation temperature of the reactor enhanced the hydrolytic activity in pig slurry and increased the volatile fatty acids (VFA) production. The short HRT (24 h) and the hyper-thermophilic temperature applied in the reactor were enough to prevent methanogenesis. No pre-treatment methods or other control methods for preventing methanogenesis were necessary. Hyper-thermophilic hydrogen production was demonstrated for the first time in a CSTR system, fed with pig slurry, using mixed culture. The results indicate that this system is a promising one for biohydrogen production from pig slurry.

Homogeneous Slowpoke reactor for the production of radio-isotope: a feasibility study

Energy Technology Data Exchange (ETDEWEB)

Busetta, P.; Bonin, H.W. [Royal Military College of Canada, Kingston, Ontario (Canada)

2006-09-15

The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP Monte Carlo reactor calculation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous react will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether natural convection can still effectively cool the reactor using the modeling software FEMLAB(r). It was found that it is needed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)

Productivity of a nuclear chemical reactor with gamma radioisotopic sources

International Nuclear Information System (INIS)

Anguis T, C.

1975-01-01

According to an established mathematical model of successive Compton interaction processes the made calculations for major distances are extended checking the acceptability of the spheric geometry model for the experimental data for radioisotopic sources of Co-60 and Cs-137. Parameters such as the increasing factor and the absorbed dose served as comparative base. calculations for the case of a punctual source succession inside a determined volume cylinder are made to obtain the total dose, the deposited energy by each photons energetic group and the total absorbed energy inside the reactor. Varying adequately the height/radius relation for different cylinders, the distinct energy depositions are compared in each one of them once a time standardized toward a standard value of energy emitted by the reactor volume. A relation between the quantity of deposited energy in each point of the reactor and the conversion values of chemical species is established. They are induced by electromagnetic radiation and that are reported as ''G'' in the scientific literature (number of molecules formed or disappeared by each 100 e.v. of energy). Once obtained the molecular performance inside the reactor for each type of geometry, it is optimized the height/radius relation according to the maximum production of molecules by unity of time. It is completed a bibliographical review of ''G'' values reported by different types of aqueous solutions with the purpose to determine the maximum performance of molecular hydrogen as a function of pH of the solution and of the used type of solute among other factors. Calculations for the ethyl bromide production as an example of one of the industrial processes which actually work using the gamma radiation as reactions inductor are realized. (Author)

Management of Hanford Site non-defense production reactor spent nuclear fuel, Hanford Site, Richland, Washington

International Nuclear Information System (INIS)

1997-03-01

The US Department of Energy (DOE) needs to provide radiologically, and industrially safe and cost-effective management of the non-defense production reactor spent nuclear fuel (SNF) at the Hanford Site. The proposed action would place the Hanford Site's non-defense production reactor SNF in a radiologically- and industrially-safe, and passive storage condition pending final disposition. The proposed action would also reduce operational costs associated with storage of the non-defense production reactor SNF through consolidation of the SNF and through use of passive rather than active storage systems. Environmental, safety and health vulnerabilities associated with existing non-defense production reactor SNF storage facilities have been identified. DOE has determined that additional activities are required to consolidate non-defense production reactor SNF management activities at the Hanford Site, including cost-effective and safe interim storage, prior to final disposition, to enable deactivation of facilities where the SNF is now stored. Cost-effectiveness would be realized: through reduced operational costs associated with passive rather than active storage systems; removal of SNF from areas undergoing deactivation as part of the Hanford Site remediation effort; and eliminating the need to duplicate future transloading facilities at the 200 and 400 Areas. Radiologically- and industrially-safe storage would be enhanced through: (1) removal from aging facilities requiring substantial upgrades to continue safe storage; (2) utilization of passive rather than active storage systems for SNF; and (3) removal of SNF from some storage containers which have a limited remaining design life. No substantial increase in Hanford Site environmental impacts would be expected from the proposed action. Environmental impacts from postulated accident scenarios also were evaluated, and indicated that the risks associated with the proposed action would be small

Refurbish research and test reactors corresponding to global age of nuclear energy

International Nuclear Information System (INIS)

Mishima, Kaichiro; Oyama, Yukio; Okamoto, Koji; Yamana, Hajime; Yamaguchi, Akira

2011-01-01

This special article featured arguments for refurbishment of research and test reactors corresponding to global age of nuclear energy, based on the report: 'Investigation of research facilities necessary for future joint usage' issued by the special committee of Atomic Energy Society of Japan (AESJ) in September 2010. It consisted of six papers titled as 'Introduction-establishment of AESJ special committee for investigation', 'State of research and test reactors in Japan', 'State of overseas research and test reactors', 'Needs analysis for research and test reactors', 'Proposal of AESJ special committee' and 'Summary and future issues'. In order to develop human resources and promote research and development needed in global age of nuclear energy, research and test reactors would be refurbished as an Asian regional center of excellence. (T. Tanaka)

Fission product release from TRIGA-LEU reactor fuels

International Nuclear Information System (INIS)

Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

1980-01-01

Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

Westinghouse independent safety review of Savannah River production reactors

International Nuclear Information System (INIS)

Leggett, W.D.; McShane, W.J.; Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E.; Call, D.W.

1989-01-01

Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K, L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours ampersand Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours ampersand Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone. 37 refs., 1 fig., 3 tabs

Westinghouse independent safety review of Savannah River production reactors

Energy Technology Data Exchange (ETDEWEB)

Leggett, W.D.; McShane, W.J. (Westinghouse Hanford Co., Richland, WA (USA)); Liparulo, N.J.; McAdoo, J.D.; Strawbridge, L.E. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear and Advanced Technology Div.); Toto, G. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Nuclear Services Div.); Fauske, H.K. (Fauske and Associates, Inc., Burr Ridge, IL (USA)); Call, D.W. (Westinghouse Savannah R

1989-04-01

Westinghouse Electric Corporation has performed a safety assessment of the Savannah River production reactors (K,L, and P) as requested by the US Department of Energy. This assessment was performed between November 1, 1988, and April 1, 1989, under the transition contract for the Westinghouse Savannah River Company's preparations to succeed E.I. du Pont de Nemours Company as the US Department of Energy contractor for the Savannah River Project. The reviewers were drawn from several Westinghouse nuclear energy organizations, embody a combination of commercial and government reactor experience, and have backgrounds covering the range of technologies relevant to assessing nuclear safety. The report presents the rationale from which the overall judgment was drawn and the basis for the committee's opinion on the phased restart strategy proposed by E.I. du Pont de Nemours Company, Westinghouse, and the US Department of Energy-Savannah River. The committee concluded that it could recommend restart of one reactor at partial power upon completion of a list of recommended upgrades both to systems and their supporting analyses and after demonstration that the organization had assimilated the massive changes it will have undergone.

The story of fission reactors: from Chicago Pile to advanced energy systems

International Nuclear Information System (INIS)

Kannan, Umasankari

2017-01-01

Nuclear reactors have been designed which cater to different applications from small research reactors of a few watts to power reactors of several Giga Watts. Based on the neutron energy, there are thermal, intermediate and fast reactors operating are being designed. On the fuel utilization front, there are designs ranging from reactors using natural uranium fuel to enriched uranium to more efficient thorium based reactors. Reactors have also been designed which are neutron eaters, minor actinide burners and breeders. There have been variety of coolant and moderating materials used for different applications from water, gas cooled, liquid sodium cooled to molten salt cooled reactors. Several new reactor designs have been developed using innovative concepts in high temperature reactors, nuclear power packs and compact reactors for special purposes. The design challenges are many from modest designs to complicated hybrid reactors. The GEN-IV forum of IAEA has selected a few of these reactor designs for commercial power production in the coming years based on several quantified indicators. The evolutionary and revolutionary design approaches have been made over the years catering to different need of energy generation. A glimpse of some of the reactors being currently developed and the design modifications done in existing reactors have been given in this paper

Medium-sized water reactors for undeveloped regions

International Nuclear Information System (INIS)

Osmachkin, V. S.

2004-01-01

In the new century the growth of population and an increasing of energy demands together with the difficulties of fossil fuel supply are expected. It is important to find optimal ways in solving such problems without the climate warming. The nuclear power having many advantages in comparison with fossil fuel technologies could play the great role in near future. The Medium-Sized Nuclear Reactors for production of electricity, heat and fresh water are considered as a main direction of nuclear power applications in the developing world It is important to discuss the requirements to such nuclear plants for using in the Countries with Small and Medium Electricity Grids. Particularly, cost-benefit analysis of construction NPP has to include assessment of all type risks and effectiveness of plant. In the paper an attention is paid on Water Reactors designed on the basis of navy technology. Such compact PWR built on special mills and placed on special floating vessel could be used in undeveloped regions. Total plant can be transported to any point of World Ocean and return back to mill for repair or decommissioning after exhaustion of lifetime. It is expected that such reactors with innovative design approach, provision of high safety and proper economic efficiency, based on leasing procedures, could be very attractive for medium-sized and developing countries.(author)

Release of radioactive fission products from BN-600 reactor untight fuel elements

International Nuclear Information System (INIS)

Osipov, S.L.; Tsikunov, A.G.; Lisitsin, E.C.

1996-01-01

The experimental data on the release of radioactive fission products from BN-600 reactor untight fuel elements are given in the report. Various groups of radionuclides: inert gases Xe, Kr, volatile Cs, J, non-volatile Nb, and La are considered. The results of calculation-experimental study of transfer and distribution of radionuclides in the reactor primary circuit, gas system and sodium coolant are considered. It is shown that some complex radioactivity transfer processes can be described by simple mathematical models. (author)

Dual-mode cultivation of Chlorella protothecoides applying inter-reactors gas transfer improves microalgae biodiesel production.

Science.gov (United States)

Santos, C A; Nobre, B; Lopes da Silva, T; Pinheiro, H M; Reis, A

2014-08-20

Chlorella protothecoides, a lipid-producing microalga, was grown heterotrophically and autotrophically in separate reactors, the off-gases exiting the former being used to aerate the latter. Autotrophic biomass productivity with the two-reactor association, 0.0249gL(-1)h(-1), was 2.2-fold the value obtained in a control autotrophic culture, aerated with ambient air. Fatty acid productivity was 1.7-fold the control value. C. protothecoides heterotrophic biomass productivity was 0.229gL(-1)h(-1). This biomass' fatty acid content was 34.5% (w/w) with a profile suitable for biodiesel production, according to European Standards. The carbon dioxide fixed by the autotrophic biomass was 45mgCO2L(-1)h(-1) in the symbiotic arrangement, 2.1 times the control reactor value. The avoided CO2 atmospheric emission represented 30% of the CO2 produced in the heterotrophic stage, while the released O2 represented 49% of the oxygen demand in that stage. Thus, an increased efficiency in the glucose carbon source use and a higher environmental sustainability were achieved in microalgal biodiesel production using the proposed assembly. Copyright © 2014 Elsevier B.V. All rights reserved.

Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

Science.gov (United States)

Sibille, Laurent; Dominques, Jesus A.

2012-01-01

The maturation of Molten Regolith Electrolysis (MRE) as a viable technology for oxygen and metals production on explored planets relies on the realization of the self-heating mode for the reactor. Joule heat generated during regolith electrolysis creates thermal energy that should be able to maintain the molten phase (similar to electrolytic Hall-Heroult process for aluminum production). Self-heating via Joule heating offers many advantages: (1) The regolith itself is the crucible material, it protects the vessel walls (2) Simplifies the engineering of the reactor (3) Reduces power consumption (no external heating) (4) Extends the longevity of the reactor. Predictive modeling is a tool chosen to perform dimensional analysis of a self-heating reactor: (1) Multiphysics modeling (COMSOL) was selected for Joule heat generation and heat transfer (2) Objective is to identify critical dimensions for first reactor prototype.

RA reactor operation and maintenance, Annual report 1974

International Nuclear Information System (INIS)

Milosevic, D. et al

1974-12-01

During 1974, RA reactor was operated at nominal power for 194 days and 13 days at lower power levels. The total production was 30711 MWh which is 2.4% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 437 users. This report contains detailed data about reactor power and experiments performed in 1974. Total number of afety shutdowns was 11, of which 8 were caused by power cuts, and 3 due to human error. Maximum individual personnel exposure dose was 50% of the maximum permissible dose. There were no accidents during this year. Decontamination of surfaces was less than during previous years. About 805 m 2 of surfaces and 178 objects were decontaminated. It was concluded that the successful operation in 1974 has a special significance taking into account the financial problems [sr

Uranium-236 in light water reactor spent fuel recycled to an enriching plant

International Nuclear Information System (INIS)

de la Garza, A.

1977-01-01

The introduction of 236 U to an enriching plant by recycling spent fuel uranium results in enriched products containing 236 U, a parasitic neutron absorber in reactor fuel. Convenient approximate methodology determines 235 236 U, and total uranium flowsheets with associated separative work requirements in enriching plant operations for use by investigators of the light water reactor fuel cycle not having recourse to specialized multicomponent cascade technology. Application of the methodology has been made to compensation of an enriching plant product for 236 U content and to the value at an enriching plant of spent fuel uranium. The approximate methodology was also confirmed with more exact calculations and with some experience with 236 U in an enriching plant

Microwave-ultrasound combined reactor suitable for atmospheric sample preparation procedure of biological and chemical products

NARCIS (Netherlands)

Lagha, A.; Chemat, S.; Bartels, P.V.; Chemat, F.

1999-01-01

A compact apparatus in which a specific position can be irradiated by microwaves (MW) and ultrasound (US) simultaneously has been developed. The MW-US reactor has been designed for atmospheric pressure digestion and dissolution of biological and chemical products. The reactor can treat a range of

RA Research reactor, Annual report 1971

International Nuclear Information System (INIS)

Milosevic, D. et al.

1971-12-01

During 1971, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and 50 days at lower power levels. Total production mounted to 31606 MWh which is 5.3% higher than planned, and the highest annual level since the reactor started operation. Reactor was used for irradiation and experiments according to the demand of 425 users, of which 370 from the Institute and 55 external users. This report contains detailed data about reactor power and experiments performed in 1971. Discrepancies from the action plan, meaning higher production was achieved in June and December due to special demand of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to power cuts during reactor operation. There was no longer interruption caused by failures of the equipment. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. Shorter interruptions resulted from breaking of connectors in the technical water pipe system caused by soil sliding near the pumping station on the Danube. Total personnel exposure dose was lower than during previous years. There was no accident nor any event that could be called accidental. Decontamination od surfaces was less than during previous years. It was concluded that the successful operation in 1971 resulted from efficient work during past years. But, some of the activities were interrupted due to undefined policy concerned with operation of the RA reactor and financial issues. This involves study of the possibility to use highly enriched fuel that would increase the useful neutron flux and the reactor compatibility with similar reactors for the future ten years. Another project that has been interrupted is construction of the emergency core cooling system which is important for the reactor safety. Financial problems have influenced not only the reactor operation but the number of employees, which could cause negative

Main problems of increasing labour productivity in the power plant construction

International Nuclear Information System (INIS)

Falaleev, P.P.

1984-01-01

The reserve for labour productivity growth in power-, industrial- and civil engineering in the USSR Minenergo system are discussed. Such reserve comprises: introduction of effective designs, increase of technological readiness of structures; a higher mechanization level in construction, improvement of industrial organization, economical and social aspects. Decrease of labour inputs in NPP construction will be attained by using unified designs of serial WWER-1000, RBMK-1000- and RBMK-1500 reactors as well as by developing nuclear power construction complexes-industrial-construction enterprises for manufacturing and transport of special structures as well as for performing civil engineering and installation work on the ground part of the reactor building and special structure. Other potentialities for increasing labour productivity in NPP construction are considered

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

Some particular aspects of control in nuclear power reactors

International Nuclear Information System (INIS)

Vathaire, F. de; Vernier, Ph.; Pascouet, A.

1964-01-01

This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [fr

Effect of furfural on ethanol production by S. cerevisiae in a cross-linked immobilized cell reactor

Energy Technology Data Exchange (ETDEWEB)

Boyer, L.J.; Vega, J.L.; Basu, R.; Clausen, E.C.; Gaddy, J.L. (Arkansas Univ., Fayetteville, AR (United States). Dept. of Chemical Engineering)

1992-01-01

Furfural, a browning reaction product, inhibits yeast (Saccharomyces cerevisiae) growth and metabolism at low concentration levels in batch culture. The performance of an immobilized cell reactor (ICR) in the presence of 0-2.0 g l[sup -1] of furfural was examined. Cell growth in the ICR, with and without furfural in the media, indicated that either furfural did not impair glucose utilization, or that the negative effects of furfural were negated by increasing cell density in the reactor. Ethanol yields were constant at 0.48 g ethanol per g glucose regardless of the furfural concentration in the media. Although the specific productivity in the ICR decreased with furfural concentration, the productivity based on liquid hold-up remained constant. Furfural was depleted in the ICR during the experimental operation. Thus, furfural levels of 2.0 g 1[sup -1] or less can be tolerated by the yeast for ethanol production in the ICR without negatively affecting reactor performance. (author).

Thermoradiation treatment of sewage sludge using reactor waste fission products

International Nuclear Information System (INIS)

Reynolds, M.C.; Hagengruber, R.L.; Zuppero, A.C.

1974-06-01

The hazards to public health associated with the application of municipal sewage sludge to land usage are reviewed to establish the need for disinfection of sludge prior to its distribution as a fertilizer, especially in the production of food and fodder. The use of ionizing radiation in conjunction with mild heating is shown to be an effective disinfection treatment and an economical one when reactor waste fission products are utilized. A program for researching and experimental demonstration of the process on sludges is also outlined

Safety analysis of an irradiation device for 99Mo production in RA-3 reactor

International Nuclear Information System (INIS)

Lerner, Ana Maria; Madariaga, Marcelo; Waldman, Ricardo

2000-01-01

The Argentine RA-3 research reactor (5 MW) has been converted to LEU fuel more than nine years ago. Since then, it has been operating with LEU fuel, which has been designed and fabricated at the National Atomic Energy Commission (CNEA). The Nuclear Regulatory Authority (ARN) is the institution in charge of the installation safety control. It is under this framework that the ARN has elaborated a neutronic calculation model for the RA-3 core, paying special attention to the device presently used for the irradiation of (HEU) 235 U targets required to obtain 9 '9Mo as a fission product. A regulatory analysis of results is carried out in the framework of ARN standards for fixed experiments. For such purpose, calculated reactivity values associated with such device are compared with recently measured values at the installation. Finally, and according to guidelines established in the first part of this work, a calculation model for a new device proposed by CNEA for the irradiation of metallic (LEU) uranium targets and still at its design stage, is here analysed. (author)

Kinetic study on the effect of temperature on biogas production using a lab scale batch reactor.

Science.gov (United States)

Deepanraj, B; Sivasubramanian, V; Jayaraj, S

2015-11-01

In the present study, biogas production from food waste through anaerobic digestion was carried out in a 2l laboratory-scale batch reactor operating at different temperatures with a hydraulic retention time of 30 days. The reactors were operated with a solid concentration of 7.5% of total solids and pH 7. The food wastes used in this experiment were subjected to characterization studies before and after digestion. Modified Gompertz model and Logistic model were used for kinetic study of biogas production. The kinetic parameters, biogas yield potential of the substrate (B), the maximum biogas production rate (Rb) and the duration of lag phase (λ), coefficient of determination (R(2)) and root mean square error (RMSE) were estimated in each case. The effect of temperature on biogas production was evaluated experimentally and compared with the results of kinetic study. The results demonstrated that the reactor with operating temperature of 50°C achieved maximum cumulative biogas production of 7556ml with better biodegradation efficiency. Copyright © 2015 Elsevier Inc. All rights reserved.

Homogeneous slowpoke reactor for the production of radio-isotope. A feasibility study

International Nuclear Information System (INIS)

Busatta, P.; Bonin, H.

2005-01-01

The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP 5 simulation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether the natural convection will still effectively cool the reactor using the modeling software FEMLAB. The MCNP 5 simulation code was validated by using a simulation with WIMS-AECL code. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)

Homogeneous slowpoke reactor for the production of radio-isotope. A feasibility study

Energy Technology Data Exchange (ETDEWEB)

Busatta, P.; Bonin, H. [Royal Military College of Canada, Kingston, Ontario (Canada)]. E-mail: paul.busatta@rmc.ca; bonin-h@rmc.ca

2005-07-01

The purpose of this research is to study the feasibility of replacing the actual heterogeneous fuel core of the present SLOWPOKE-2 by a reservoir containing a homogeneous fuel for the production of Mo-99. The study looked at three items: by using the MCNP 5 simulation code, develop a series of parameters required for an homogeneous fuel and evaluate the uranyl sulfate concentration of the aqueous solution fuel in order to keep a similar excess reactivity; verify if the homogeneous reactor will retain its inherent safety attributes; and with the new dimensions and geometry of the fuel core, observe whether the natural convection will still effectively cool the reactor using the modeling software FEMLAB. The MCNP 5 simulation code was validated by using a simulation with WIMS-AECL code. It was found that it is indeed feasible to modify the SLOWPOKE-2 reactor for a homogeneous reactor using a solution of uranyl sulfate and water. (author)

CFD analysis and flow model reduction for surfactant production in helix reactor

Directory of Open Access Journals (Sweden)

Nikačević N.M.

2015-01-01

Full Text Available Flow pattern analysis in a spiral Helix reactor is conducted, for the application in the commercial surfactant production. Step change response curves (SCR were obtained from numerical tracer experiments by three-dimensional computational fluid dynamics (CFD simulations. Non-reactive flow is simulated, though viscosity is treated as variable in the direction of flow, as it increases during the reaction. The design and operating parameters (reactor diameter, number of coils and inlet velocity are varied in CFD simulations, in order to examine the effects on the flow pattern. Given that 3D simulations are not practical for fast computations needed for optimization, scale-up and control, CFD flow model is reduced to one-dimensional axial dispersion (AD model with spatially variable dispersion coefficient. Dimensionless dispersion coefficient (Pe is estimated under different conditions and results are analyzed. Finally, correlation which relates Pe number with Reynolds number and number of coils from the reactor entrance is proposed for the particular reactor application and conditions.

Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor

Science.gov (United States)

2012-01-01

Background A modified laboratory-scale upflow anaerobic sludge blanket (UASB) reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in agriculture without the risk of polluting soil, underground water or crops. In this sense, treatment of vinasse combines the elimination of organic waste with the formation of methane. Biogas is considered as a promising renewable energy source. The aim of this study was to determine the optimum organic loading rate for operating a modified UASB reactor to treat vinasse generated in the production of hydrous ethanol from sugar cane molasses. Results The study showed that chemical oxygen demand (COD) removal efficiency was 69% at an optimum organic loading rate (OLR) of 17.05 kg COD/m3-day, achieving a methane yield of 0.263 m3/kg CODadded and a biogas methane content of 84%. During this stage, effluent characterization presented lower values than the vinasse, except for potassium, sulfide and ammonia nitrogen. On the other hand, primers used to amplify the 16S-rDNA genes for the domains Archaea and Bacteria showed the presence of microorganisms which favor methane production at the optimum organic loading rate. Conclusions The modified UASB reactor proposed in this study provided a successful treatment of the vinasse obtained from hydrous ethanol production. Methanogen groups (Methanobacteriales and Methanosarcinales) detected by PCR during operational optimum OLR of the modified UASB reactor, favored methane production. PMID:23167984

Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor

Directory of Open Access Journals (Sweden)

España-Gamboa Elda I

2012-11-01

Full Text Available Abstract Background A modified laboratory-scale upflow anaerobic sludge blanket (UASB reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in agriculture without the risk of polluting soil, underground water or crops. In this sense, treatment of vinasse combines the elimination of organic waste with the formation of methane. Biogas is considered as a promising renewable energy source. The aim of this study was to determine the optimum organic loading rate for operating a modified UASB reactor to treat vinasse generated in the production of hydrous ethanol from sugar cane molasses. Results The study showed that chemical oxygen demand (COD removal efficiency was 69% at an optimum organic loading rate (OLR of 17.05 kg COD/m3-day, achieving a methane yield of 0.263 m3/kg CODadded and a biogas methane content of 84%. During this stage, effluent characterization presented lower values than the vinasse, except for potassium, sulfide and ammonia nitrogen. On the other hand, primers used to amplify the 16S-rDNA genes for the domains Archaea and Bacteria showed the presence of microorganisms which favor methane production at the optimum organic loading rate. Conclusions The modified UASB reactor proposed in this study provided a successful treatment of the vinasse obtained from hydrous ethanol production. Methanogen groups (Methanobacteriales and Methanosarcinales detected by PCR during operational optimum OLR of the modified UASB reactor, favored methane production.

Methane production by treating vinasses from hydrous ethanol using a modified UASB reactor.

Science.gov (United States)

España-Gamboa, Elda I; Mijangos-Cortés, Javier O; Hernández-Zárate, Galdy; Maldonado, Jorge A Domínguez; Alzate-Gaviria, Liliana M

2012-11-21

A modified laboratory-scale upflow anaerobic sludge blanket (UASB) reactor was used to obtain methane by treating hydrous ethanol vinasse. Vinasses or stillage are waste materials with high organic loads, and a complex composition resulting from the process of alcohol distillation. They must initially be treated with anaerobic processes due to their high organic loads. Vinasses can be considered multipurpose waste for energy recovery and once treated they can be used in agriculture without the risk of polluting soil, underground water or crops. In this sense, treatment of vinasse combines the elimination of organic waste with the formation of methane. Biogas is considered as a promising renewable energy source. The aim of this study was to determine the optimum organic loading rate for operating a modified UASB reactor to treat vinasse generated in the production of hydrous ethanol from sugar cane molasses. The study showed that chemical oxygen demand (COD) removal efficiency was 69% at an optimum organic loading rate (OLR) of 17.05 kg COD/m3-day, achieving a methane yield of 0.263 m3/kg CODadded and a biogas methane content of 84%. During this stage, effluent characterization presented lower values than the vinasse, except for potassium, sulfide and ammonia nitrogen. On the other hand, primers used to amplify the 16S-rDNA genes for the domains Archaea and Bacteria showed the presence of microorganisms which favor methane production at the optimum organic loading rate. The modified UASB reactor proposed in this study provided a successful treatment of the vinasse obtained from hydrous ethanol production.Methanogen groups (Methanobacteriales and Methanosarcinales) detected by PCR during operational optimum OLR of the modified UASB reactor, favored methane production.

Factors governing particulate corrosion product adhesion to surfaces in water reactor coolant circuits

International Nuclear Information System (INIS)

1979-03-01

Gravity, van der Waals, magnetic, electrical double layer and hydrodynamic forces are considered as potential contributors to the adhesion of particulate corrosion products to surfaces in water reactor coolant circuits. These forces are renewed and evaluated, and the following are amongst the conclusions drawn; adequate theories are available to estimate the forces governing corrosion product particle adhesion to surfaces in single phase flow in water reactor coolant circuits. Some uncertainty is introduced by the geometry of real particle-surface systems. The major uncertainties are due to inadequate data on the Hamaker constant and the zeta potential for the relevant materials, water chemistry and radiation chemistry at 300 0 C; van der Waals force is dominant over the effect of gravity for particles smaller than about 100 m; quite modest zeta potentials, approximately 50mV, are capable of inhibiting particle deposition throughout the size range relevant to water reactors; for surfaces exposed to typical water reactor flow conditions, particles smaller than approximately 1 m will be stable against resuspension in the absence of electrical double layer repulsion; and the magnitude of the electrical double layer repulsion for a given potential depends on whether the interaction is assumed to occur at constant potential or constant change. (author)

A highly efficient autothermal microchannel reactor for ammonia decomposition: Analysis of hydrogen production in transient and steady-state regimes

Science.gov (United States)

Engelbrecht, Nicolaas; Chiuta, Steven; Bessarabov, Dmitri G.

2018-05-01

The experimental evaluation of an autothermal microchannel reactor for H2 production from NH3 decomposition is described. The reactor design incorporates an autothermal approach, with added NH3 oxidation, for coupled heat supply to the endothermic decomposition reaction. An alternating catalytic plate arrangement is used to accomplish this thermal coupling in a cocurrent flow strategy. Detailed analysis of the transient operating regime associated with reactor start-up and steady-state results is presented. The effects of operating parameters on reactor performance are investigated, specifically, the NH3 decomposition flow rate, NH3 oxidation flow rate, and fuel-oxygen equivalence ratio. Overall, the reactor exhibits rapid response time during start-up; within 60 min, H2 production is approximately 95% of steady-state values. The recommended operating point for steady-state H2 production corresponds to an NH3 decomposition flow rate of 6 NL min-1, NH3 oxidation flow rate of 4 NL min-1, and fuel-oxygen equivalence ratio of 1.4. Under these flows, NH3 conversion of 99.8% and H2 equivalent fuel cell power output of 0.71 kWe is achieved. The reactor shows good heat utilization with a thermal efficiency of 75.9%. An efficient autothermal reactor design is therefore demonstrated, which may be upscaled to a multi-kW H2 production system for commercial implementation.

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

International Nuclear Information System (INIS)

Friess, Friederike Renate

2017-01-01

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike Renate

2017-07-12

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Introduction to the special issue on the technical status of materials for a fusion reactor

Science.gov (United States)

Stork, D.; Zinkle, S. J.

2017-09-01

Materials determine in a fundamental way the performance and environmental attractiveness of a fusion reactor: through the size (power fluxes to the divertor, neutron fluxes to the first wall); economics (replacement lifetime of critical in-vessel components, thermodynamic efficiency through operating temperature etc); plasma performance (erosion by plasma fluxes to the divertor surfaces); robustness against off-normal accidents (safety); and the effects of post-operation radioactivity on waste disposal and maintenance. The major philosophies and methodologies used to formulate programmes for the development of fusion materials are outlined, as the basis for other articles in this special issue, which deal with the fundamental understanding of the issues regarding these materials and their technical status and prospects for development.

Production of {sup 48}V in a nuclear reactor via secondary tritons

Energy Technology Data Exchange (ETDEWEB)

Siri, S. [Comision Nacional de Energia Atomica, Centro Atomico Ezeiza, Gerencia de Capacitacion, Quimica Nuclear y Ciencias de la Salud, Ezeiza, Buenos Aires (Argentina); Cohen, I.M. [Univ. Tecnologica Nacional, Dept. de Ingenieria Quimica, Buenos Aires (Argentina)

2009-07-01

The production of {sup 48}V in a nuclear reactor, induced on titanium by tritons generated from the {sup 6}Li(n, t){sup 4} He reaction, and eventually {sup 7}Li(n, n't){sup 4}He, is described. Samples of lithium titanate were irradiated for an irradiation cycle (120 h) in the RA-3 reactor, belonging to Ezeiza Atomic Centre. After a radiochemical separation, the characteristic radiations from {sup 48}V were identified in the gamma ray spectra of the vanadium fractions. (orig.)

Nuclear reactor plant for production process heat

International Nuclear Information System (INIS)

Weber, M.

1979-01-01

The high temperature reactor is suitable as a heat source for carrying out endothermal chemical processes. A heat exchanger is required for separating the reactor coolant gases and the process medium. The heat of the reactor is transferred at a temperature lower than the process temperature to a secondary gas and is compressed to give the required temperature. The compression energy is obtained from the same reactor. (RW) [de

A comprehensive review of microbial electrolysis cells (MEC reactor designs and configurations for sustainable hydrogen gas production

Directory of Open Access Journals (Sweden)

Abudukeremu Kadier

2016-03-01

Full Text Available Hydrogen gas has tremendous potential as an environmentally acceptable energy carrier for vehicles. A cutting edge technology called a microbial electrolysis cell (MEC can achieve sustainable and clean hydrogen production from a wide range of renewable biomass and wastewaters. Enhancing the hydrogen production rate and lowering the energy input are the main challenges of MEC technology. MEC reactor design is one of the crucial factors which directly influence on hydrogen and current production rate in MECs. The rector design is also a key factor to up-scaling. Traditional MEC designs incorporated membranes, but it was recently shown that membrane-free designs can lead to both high hydrogen recoveries and production rates. Since then multiple studies have developed reactors that operate without membranes. This review provides a brief overview of recent advances in research on scalable MEC reactor design and configurations.

PRELIMINARY DESIGN OF OSCILLATORY FLOW BIODIESEL REACTOR FOR CONTINUOUS BIODIESEL PRODUCTION FROM JATROPHA TRIGLYCERIDES

Directory of Open Access Journals (Sweden)

AZHARI T. I. MOHD. GHAZI

2008-08-01

Full Text Available The concept of a continuous process in producing biodiesel from jatropha oil by using an Oscillatory Flow Biodiesel Reactor (OFBR is discussed in this paper. It has been recognized that the batch stirred reactor is a primary mode used in the synthesis of biodiesel. However, pulsatile flow has been extensively researcehed and the fundamental principles have been successfully developed upon which its hydrodynamics are based. Oscillatory flow biodiesel reactor offers precise control of mixing by means of the baffle geometry and pulsation which facilitates to continuous operation, giving plug flow residence time distribution with high turbulence and enhanced mass and heat transfer. In conjunction with the concept of reactor design, parameters such as reactor dimensions, the hydrodynamic studies and physical properties of reactants must be considered prior to the design work initiated recently. The OFBR reactor design involves the use of simulation software, ASPEN PLUS and the reactor design fundamentals. Following this, the design parameters shall be applied in fabricating the OFBR for laboratory scale biodiesel production.

Production and characterization of scum and its role in odour control in UASB reactors treating domestic wastewater.

Science.gov (United States)

Souza, C L; Silva, S Q; Aquino, S F; Chernicharo, C A L

2006-01-01

There are few studies in the literature that have aimed at characterizing the physical, chemical, and microbial aspects of scum produced in UASB reactors. In addition, there is little information on the influence of operational conditions of UASB reactors on scum formation, and the present work addresses these issues. Three demo-scale UASB reactors, fed on domestic wastewater, were employed to monitor the formation and its characteristics. Scum production was periodically assessed during different operational phases, and its characterization involved analyses of BOD, COD, solids, sulfide, sulfate, microscopic observations, as well as biodegradability tests. The results show that the scum formed was physically, chemically, and microscopically similar in both geminated reactors, being comprised mainly of organic material of low biodegradability. Several bacterial morphotypes, mainly filaments and rods, with internal sulfur granules, were observed, and the aerobic microorganisms that developed at the scum layer as a result of photosynthetic activity of cyanobacteria, seemed to play an important role in sulfide removal and odour control. Scum production rates were similar in both reactors, but the imposed higher upflow velocities resulted in a higher production rate and in a reduced biodegradability of the scum.

Production and validation of nuclear data for reactor and fuel cycle applications

International Nuclear Information System (INIS)

Trakas, C.; Verwaerde, D.; Toubon, H.

2002-01-01

The aim of this technical meeting is the improvement of the existing nuclear data and the production of new data of interest for the upstream and downstream of the fuel cycle (enrichment, fabrication, management, storage, transport, reprocessing), for the industrial reactors, the research reactors and the new reactor concepts (criticality, dimensioning, exploitation), for the instrumentation systems (external and internal sensors), the radioprotection, the residual power, the structures (neutron bombardment effect on vessels, rods etc..), and for the activation of steel structures (Fr, Ni, Co). The expected result is the collection of more reliable and accurate data in a wider spectrum of energies and temperatures thanks to more precise computer codes and measurement techniques. This document brings together the communications presented at this meeting and dealing with: the process of production and validation of nuclear data; the measurement facilities and the big international programs; the users needs and the industrial priorities; the basic nuclear data (BND) needs at Cogema; the expression and evaluation of BND; the evaluation work: the efficient cross-sections; the processing of data and the creation of activation libraries; from the integral measurement to the qualification and the feedback on nuclear data. (J.S.)

A study on conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production reactor based on spherical torus (ST), which is an intermediate application of fusion energy, is presented. Different from traditional Tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST are used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can do within vacuum vessel in order to produce certain amount of excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR is presented. Based on systematical analysis, design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (authors)

Syngas fermentation by Clostridium carboxidivorans P7 in a horizontal rotating packed bed biofilm reactor with enhanced ethanol production

International Nuclear Information System (INIS)

Shen, Yanwen; Brown, Robert C.; Wen, Zhiyou

2017-01-01

Highlights: • A novel a horizontal rotating packed bed (h-RPB) reactor for syngas fermentation was reported. • The h-RPB reactor enhanced ethanol productivity by 3.3-folds compared to continuous stirred tank reactor (CSTR). • The h-RPB reactor has a unique feature of transfer gas from both bulk liquid phase and headspace phase. • The mass transfer in the headspace of h-PRB played an important role for enhanced ethanol production. - Abstract: Gasification of lignocellulosic biomass followed by syngas fermentation is a promising process for producing fuels and chemicals. Syngas fermentation, however, is commonly limited by low mass transfer rates. In this work, a horizontally oriented rotating packed bed (h-RPB) reactor was developed to improve mass transfer and enhance ethanol production. In the h-RPB reactor, cell attachment materials were packed in the reactor and half submerged in the liquid and half exposed to the headspace. With continuous rotation of the packing materials, the cells in biofilm were alternately in contact with liquid and headspace; thus, transport of syngas to the cells occurred in both the liquid phase and headspace. The volumetric mass transfer coefficient (k_La) of the h-RPB reactor was lower than that in a traditional continuous stirred tank reactor (CSTR), indicating the mass transfer in the liquid phase of h-PRB was lower than CSTR, and the mass transfer in the headspace phase played an important role in syngas fermentation. The syngas fermentation of Clostridium carboxidivorans P7 in h-RPB resulted in a 7.0 g/L titer and 6.7 g/L/day productivity of ethanol, respectively, 3.3 times higher than those obtained in a CSTR under the same operational conditions. The results demonstrate that the h-RPB reactor is an efficient system for syngas fermentation, making cellulosic ethanol biorefinery one step closer to technical and economic feasibility.

Decommissioning and dismantling reactors and managing waste

International Nuclear Information System (INIS)

Bensoussan, E.; Reicher-Fournel, N.

2005-01-01

In the early forties/fifties, a number of countries launched the first developments in the field of nuclear power. Some of them now have large numbers of nuclear facilities and nuclear power plants which have met, and continue to meet, the objectives for which they were designed and built. Other plants, including nuclear fuel production and enrichment plants, experimental reactors or research reactors, will have to be dismantled and demolished in the near future. These activities are handled differently in different countries as a function of specific energy policies, advanced development plants, current financial resources, the availability of qualified engineers and specialized industries able to handle projects of this kind, as well as other factors. All dismantling and demolition projects serve the purpose of returning the respective sites to green-field conditions. (orig.)

Vectorization and parallelization of a production reactor assembly code

International Nuclear Information System (INIS)

Vujic, J.L.; Martin, W.R.; Michigan Univ., Ann Arbor, MI

1991-01-01

In order to use efficiently the new features of supercomputers, production codes, usually written 10 -20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantial speedup in the execution times was obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize the parallel/vector architecture of a multiprocessor IBM 3090. (author)

Vectorization and parallelization of a production reactor assembly code

International Nuclear Information System (INIS)

Vujic, J.L.; Martin, W.R.

1991-01-01

In order to efficiently use new features of supercomputers, production codes, usually written 10 - 20 years ago, must be tailored for modern computer architectures. We have chosen to optimize the CPM-2 code, a production reactor assembly code based on the collision probability transport method. Substantial speedups in the execution times were obtained with the parallel/vector version of the CPM-2 code. In addition, we have developed a new transfer probability method, which removes some of the modelling limitations of the collision probability method encoded in the CPM-2 code, and can fully utilize parallel/vector architecture of a multiprocessor IBM 3090. (author)

Use of a two-chamber reactor to improve enzymatic hydrolysis and fermentation of lignocellulosic materials

International Nuclear Information System (INIS)

Viola, E.; Zimbardi, F.; Valerio, V.; Nanna, F.; Battafarano, A.

2013-01-01

Highlights: ► A two-chamber reactor is proposed to improve bioethanol production. ► Hydrolysis and fermentation can be made simultaneous at different temperatures. ► The residue of lignin can be easily separated at the end of the process. -- Abstract: A special type of bioreactor was designed and tested in order to improve the bioethanol production from lignocellulosic materials via enzymatic hydrolysis and fermentation. The reactor consists of two chambers kept at different temperatures and separated by a porous medium, through which the solutes can diffuse. The reactor was tested using as substrate wheat straw previously steam exploded and detoxified. The yields of cellulose hydrolysis and glucose fermentation obtained using this reactor were compared to those obtained by simultaneous enzymatic hydrolysis and fermentation (SSF) carried out in only one vessel. The results showed that a significant increase in the ethanol yield (20%) can be achieved by using this bioreactor. An additional advantage of the reactor is the confinement of the solid lignin in one chamber, allowing a simplified separation process between broth and unreacted residue.

Breeder reactors

International Nuclear Information System (INIS)

Gollion, H.

1977-01-01

The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

Hydrogen production from fusion reactors coupled with high temperature electrolysis

International Nuclear Information System (INIS)

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

The decreasing availability of fossil fuels emphasizes the need to develop systems which will produce synthetic fuel to substitute for and complement the natural supply. An important first step in the synthesis of liquid and gaseous fuels is the production of hydrogen. Thermonuclear fusion offers an inexhaustible source of energy for the production of hydrogen from water. Processes which may be considered for this purpose include electrolysis, thermochemical decomposition or thermochemical-electrochemical hybrid cycles. Preliminary studies at Brookhaven indicate that high temperature electrolysis has the highest potential efficiency for production of hydrogen from fusion. Depending on design electric generation efficiencies of approximately 40 to 60 percent and hydrogen production efficiencies of approximately 50 to 70 percent are projected for fusion reactors using high temperature blankets

Development of the production of special steels for nuclear industries

International Nuclear Information System (INIS)

Vieillard-Baron, B.

1977-01-01

The development of electro-nuclear industries has a powerful impact on the production of special steels, although the quantity of material applied to the non-conventional parts of nuclear power plants is quite small as compared to the total production requirements in this industrial field. Evolution bears on the product research, development and testing methods, on the technical and marketing services - in particular the establishment of quality control teams and assurance manuals - and the implementation of high performance production equipments. Manufacturing must however take place under normal work load and productivity conditions of production tools, and thus ensure a satisfactory profitability on investments entailed [fr

The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

International Nuclear Information System (INIS)

Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

1992-01-01

A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

Hydrogen production by high-temperature gas-cooled reactor. Conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system

International Nuclear Information System (INIS)

Sakaba, Nariaki; Ohashi, Hirofumi; Sato, Hiroyuki; Hara, Teruo; Kato, Ryoma; Kunitomi, Kazuhiko

2008-01-01

Nuclear hydrogen production is necessary in an anticipated hydrogen society that demands a massive quantity of hydrogen without economic disadvantage. Japan Atomic Energy Agency (JAEA) has launched the conceptual design study of a hydrogen production system with a near-term plan to connect it to Japan's first high-temperature gas-cooled reactor HTTR. The candidate hydrogen production system is based on the thermochemical water-splitting iodine sulphur (IS) process.The heat of 10 MWth at approximately 900degC, which can be provided by the secondary helium from the intermediate heat exchanger of the HTTR, is the energy input to the hydrogen production system. In this paper, we describe the recent progresses made in the conceptual design of advanced process heat exchangers of the HTTR-IS hydrogen production system. A new concept of sulphuric acid decomposer is proposed. This involves the integration of three separate functions of sulphuric acid decomposer, sulphur trioxide decomposer, and process heat exchanger. A new mixer-settler type of Bunsen reactor is also designed. This integrates three separate functions of Bunsen reactor, phase separator, and pump. The new concepts are expected to result in improved economics through construction and operation cost reductions because the number of process equipment and complicated connections between the equipment has been substantially reduced. (author)

Fusion-fission hybrid reactors

International Nuclear Information System (INIS)

Greenspan, E.

1984-01-01

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