Operational experience with Dragon reactor experiment of relevance to commercial reactors

International Nuclear Information System (INIS)

Capp, P.D.; Simon, R.A.

1976-01-01

An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

The Dragon reactor experiment

International Nuclear Information System (INIS)

Anon.

1976-01-01

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

Dynamics of soil water evaporation during soil drying: laboratory experiment and numerical analysis.

Science.gov (United States)

Han, Jiangbo; Zhou, Zhifang

2013-01-01

Laboratory and numerical experiments were conducted to investigate the evolution of soil water evaporation during a continuous drying event. Simulated soil water contents and temperatures by the calibrated model well reproduced measured values at different depths. Results show that the evaporative drying process could be divided into three stages, beginning with a relatively high evaporation rate during stage 1, followed by a lower rate during transient stage and stage 2, and finally maintaining a very low and constant rate during stage 3. The condensation zone was located immediately below the evaporation zone in the profile. Both peaks of evaporation and condensation rate increased rapidly during stage 1 and transition stage, decreased during stage 2, and maintained constant during stage 3. The width of evaporation zone kept a continuous increase during stages 1 and 2 and maintained a nearly constant value of 0.68 cm during stage 3. When the evaporation zone totally moved into the subsurface, a dry surface layer (DSL) formed above the evaporation zone at the end of stage 2. The width of DSL also presented a continuous increase during stage 2 and kept a constant value of 0.71 cm during stage 3.

Solvent refined coal reactor quench system

Science.gov (United States)

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

Energy Technology Data Exchange (ETDEWEB)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H., E-mail: songm@dgu.edu [Department of Mechanical, Robotics, and Energy Engineering, Dongguk University, Seoul 100-715 (Korea, Republic of)

2015-03-15

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made.

Tetrafluoroethane (R134a) hydrate formation within variable volume reactor accompanied by evaporation and condensation

International Nuclear Information System (INIS)

Jeong, K.; Choo, Y. S.; Hong, H. J.; Yoon, Y. S.; Song, M. H.

2015-01-01

Vast size hydrate formation reactors with fast conversion rate are required for the economic implementation of seawater desalination utilizing gas hydrate technology. The commercial target production rate is order of thousand tons of potable water per day per train. Various heat and mass transfer enhancement schemes including agitation, spraying, and bubbling have been examined to maximize the production capacities in scaled up design of hydrate formation reactors. The present experimental study focused on acquiring basic knowledge needed to design variable volume reactors to produce tetrafluoroethane hydrate slurry. Test vessel was composed of main cavity with fixed volume of 140 ml and auxiliary cavity with variable volume of 0 ∼ 64 ml. Temperatures at multiple locations within vessel and pressure were monitored while visual access was made through front window. Alternating evaporation and condensation induced by cyclic volume change provided agitation due to density differences among water and vapor, liquid and hydrate R134a as well as extended interface area, which improved hydrate formation kinetics coupled with latent heat release and absorption. Influences of coolant temperature, piston stroke/speed, and volume change period on hydrate formation kinetics were investigated. Suggestions of reactor design improvement for future experimental study are also made

Novel evaporation experiment to determine soil hydraulic properties

Directory of Open Access Journals (Sweden)

K. Schneider

2006-01-01

Full Text Available A novel experimental approach to determine soil hydraulic material properties for the dry and very dry range is presented. Evaporation from the surface of a soil column is controlled by a constant flux of preconditioned air and the resulting vapour flux is measured by infrared absorption spectroscopy. The data are inverted under the assumptions that (i the simultaneous movement of water in the liquid and vapour is represented by Richards' equation with an effective hydraulic conductivity and that (ii the coupling between the soil and the well-mixed atmosphere can be modelled by a boundary layer with a constant transfer resistance. The optimised model fits the data exceptionally well. Remaining deviations during the initial phase of an experiment are thought to be well-understood and are attributed to the onset of the heat flow through the column which compensates the latent heat of evaporation.

Dynamics of Soil Water Evaporation during Soil Drying: Laboratory Experiment and Numerical Analysis

Science.gov (United States)

Han, Jiangbo; Zhou, Zhifang

2013-01-01

Laboratory and numerical experiments were conducted to investigate the evolution of soil water evaporation during a continuous drying event. Simulated soil water contents and temperatures by the calibrated model well reproduced measured values at different depths. Results show that the evaporative drying process could be divided into three stages, beginning with a relatively high evaporation rate during stage 1, followed by a lower rate during transient stage and stage 2, and finally maintaining a very low and constant rate during stage 3. The condensation zone was located immediately below the evaporation zone in the profile. Both peaks of evaporation and condensation rate increased rapidly during stage 1 and transition stage, decreased during stage 2, and maintained constant during stage 3. The width of evaporation zone kept a continuous increase during stages 1 and 2 and maintained a nearly constant value of 0.68 cm during stage 3. When the evaporation zone totally moved into the subsurface, a dry surface layer (DSL) formed above the evaporation zone at the end of stage 2. The width of DSL also presented a continuous increase during stage 2 and kept a constant value of 0.71 cm during stage 3. PMID:24489492

Simultaneous ion and neutral evaporation in aqueous nanodrops: experiment, theory, and molecular dynamics simulations.

Science.gov (United States)

Higashi, Hidenori; Tokumi, Takuya; Hogan, Christopher J; Suda, Hiroshi; Seto, Takafumi; Otani, Yoshio

2015-06-28

We use a combination of tandem ion mobility spectrometry (IMS-IMS, with differential mobility analyzers), molecular dynamics (MD) simulations, and analytical models to examine both neutral solvent (H2O) and ion (solvated Na(+)) evaporation from aqueous sodium chloride nanodrops. For experiments, nanodrops were produced via electrospray ionization (ESI) of an aqueous sodium chloride solution. Two nanodrops were examined in MD simulations: a 2500 water molecule nanodrop with 68 Na(+) and 60 Cl(-) ions (an initial net charge of z = +8), and (2) a 1000 water molecule nanodrop with 65 Na(+) and 60 Cl(-) ions (an initial net charge of z = +5). Specifically, we used MD simulations to examine the validity of a model for the neutral evaporation rate incorporating both the Kelvin (surface curvature) and Thomson (electrostatic) influences, while both MD simulations and experimental measurements were compared to predictions of the ion evaporation rate equation of Labowsky et al. [Anal. Chim. Acta, 2000, 406, 105-118]. Within a single fit parameter, we find excellent agreement between simulated and modeled neutral evaporation rates for nanodrops with solute volume fractions below 0.30. Similarly, MD simulation inferred ion evaporation rates are in excellent agreement with predictions based on the Labowsky et al. equation. Measurements of the sizes and charge states of ESI generated NaCl clusters suggest that the charge states of these clusters are governed by ion evaporation, however, ion evaporation appears to have occurred with lower activation energies in experiments than was anticipated based on analytical calculations as well as MD simulations. Several possible reasons for this discrepancy are discussed.

The experience of liquid radwaste evaporator performance improvement

International Nuclear Information System (INIS)

Kwon, S. H.

1997-01-01

Ulchin NPP has only one monobloc evaporation column which treated all radwaste liquid for two units. Since commercial operation in 1988 the evaporator performance is very poor. I think that the bad condition of evaporator is because of the bad quality of liquid radwaste, the large volume of liquid radwaste to treated, the poor skill of operation and some mistake in equipment design. Because of above conditions the average released activity by liquid radwaste is 35.153mCi/year in last eight years(1988∼1995). So it is necessary that we have to improve the evaporator performance and to reduce the liquid radwaste volume to evaporate

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

Energy Technology Data Exchange (ETDEWEB)

Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

2017-07-01

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

Study on reactor power transient characteristics (reactor training experiments). Control rod reactivity calibration by positive period method and other experiment

International Nuclear Information System (INIS)

Ozaki, Yoshihiko; Sunagawa, Takeyoshi

2014-01-01

In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity calibration experiments using positive Period method were carried out for shim safety rod and regulating rod, respectively. The results were obtained as reasonable values in comparison with the nominal value of the UTR-KINKI. The measurement of reactor power change after reactor scram was performed, and the presence of the delayed neutron precursor was confirmed by calculating the half-life. The spatial dose rate measurement experiment of neutrons and γ-rays in the reactor room in a reactor power 1W operating conditions were also performed. (author)

Effect of water losses by evaporation and chemical reaction in an industrial slaker reactor

Directory of Open Access Journals (Sweden)

Ricardo Andreola

2007-03-01

Full Text Available A dynamic model of the slaker reactor was developed and validated for Klabin Paraná Papéis causticizing system, responsable for white liquor generation used by the plant. The model considered water losses by evaporation and chemical reaction. The model showed a good agreement with the industrial plant measures of active alkali, total titratable alkali and temperature, without the need of adjustment of any parameter. The simulated results showed that the water consumption by the slaking reaction and evaporation exerted significant influence on the volumetric flow rate of limed liquor, which imposed a decrease of 4.6% in the amount of water in reactor outlet.Foi desenvolvido e testado um modelo dinâmico do reator de apagamento do sistema de caustificação da Klabin Paraná Papéis, responsável pela geração do licor branco utilizado na planta. O modelo contempla perdas de água por evaporação e por reação química e apresentou boa concordância com dados industriais de álcali ativo, álcali total titulável e temperatura, sem a necessidade de ajuste de nenhum parâmetro. Os resultados obtidos a partir de simulações revelam que o consumo de água pela reação de apagamento, bem como pela evaporação, exercem uma influência significativa sobre a vazão volumétrica na saída do reator, impondo uma diminuição de 4,6% sobre o teor de água na corrente de saída do reator em relação à alimentação.

Impact of an Ultraviolet Reactor on the Improvement of Air Quality Leaving a Direct Evaporative Cooler

Directory of Open Access Journals (Sweden)

Wonjun Kim

2018-04-01

Full Text Available The purpose of this study is to improve microbial air quality by improving water quality, particularly concerning microbiological aspects, by applying an ultraviolet water purifier system to a direct evaporative cooling (DEC system. A direct evaporative cooler is an air cooling technique that uses the evaporation of water. Most DECs recirculate water to reduce water use. Evaporative cooling pads and water are biologically contaminated by recirculating water. This contamination can develop into air contamination and cause respiratory illnesses in occupants. It is necessary to use sterilized water in a DEC to prevent respiratory diseases and maintain air quality. In this study, we examine whether improvements in water quality in a DEC affect air quality by dividing experiments into a control group (Control and a treated group (UV-treated. In the control group, the degree of contamination was measured when a DEC operated for four weeks without ultraviolet water treatment. In UV-treated, the degree of contamination was measured when UV water treatment was applied to a DEC for four weeks. In both Control and UV-treated, microbes were sampled from the water, the evaporative cooling pad surface, and the DEC inlet and outlet air samples in order to compare the levels of contamination. The surface was measured once at four points, and the air was measured four times at two points. A comparison of the two experiments indicated that the degree of microbial contamination of water and air was significantly reduced in the UV-treated group when compared to that in the control group. When the pollution degree of the evaporative cooling pad was compared to the degree of air pollution, it was difficult to obtain a correlation between the two factors, although the results confirmed that the contamination of the evaporative cooling pad caused water pollution. Therefore, it is necessary to operate a water treatment system to maintain the clean air in DECs.

Evaporation experiments and modelling for glass melts

NARCIS (Netherlands)

Limpt, J.A.C. van; Beerkens, R.G.C.

2007-01-01

A laboratory test facility has been developed to measure evaporation rates of different volatile components from commercial and model glass compositions. In the set-up the furnace atmosphere, temperature level, gas velocity and batch composition are controlled. Evaporation rates have been measured

Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

International Nuclear Information System (INIS)

Yamada, Fumiaki; Mori, Takero

2005-01-01

In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Results of the Nucifer reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Buck, Christian; Lindner, Manfred [MPIK Heidelberg (Germany)

2016-07-01

Nuclear reactors are a strong and pure source of electron antineutrinos. With neutrino experiments close to compact reactor cores new insights into neutrino properties and reactor physics can be obtained. The Nucifer experiment is one of the pioneers in this class of very short baseline projects. Its detector to reactor distance is only about 7 m. The data obtained in the last years allowed to estimate the plutonium concentration in the reactor core by the neutrino flux measurement. This is of interest for safeguard applications and non proliferation efforts. The antineutrinos in Nucifer are detected via the inverse beta decay on free protons. Those Hydrogen nuclei are provided by 850 liters of organic liquid scintillator. For higher detection efficiency and background reduction the liquid is loaded with Gadolinium. Despite all shielding efforts and veto systems the background induced by the reactor activity and cosmogenic particles is still the main challenge in the experiment. The principle of the Nucifer detector is similar to the needs of upcoming experiments searching for sterile neutrinos. Therefore, the Nucifer results are also valuable input for the understanding and optimization of those next generation projects. The observation of sterile neutrinos would imply new physics beyond the standard model.

[Dynamics of Irreversible Evaporation of a Water-Protein Droplet and a Problem of Structural and Dynamical Experiments with Single Molecules].

Science.gov (United States)

Shaitan, K V; Armeev, G A; Shaytan, A K

2016-01-01

We discuss the effect of isothermal and adiabatic evaporation of water on the state of a water-protein droplet. The discussed problem is of current importance due to development of techniques to perform single molecule experiments using free electron lasers. In such structure-dynamic experiments the delivery of a sample into the X-ray beam is performed using the microdroplet injector. The time between the injection and delivery is in the order of microseconds. In this paper we developed a specialized variant of all-atom molecular dynamics simulations for the study of irreversible isothermal evaporation of the droplet. Using in silico experiments we determined the parameters of isothermal evaporation of the water-protein droplet with the sodium and chloride ions in the concentration range of 0.3 M at different temperatures. The energy of irreversible evaporation determined from in silico experiments at the initial stages of evaporation virtually coincides with the specific heat of evaporation for water. For the kinetics of irreversible adiabatic evaporation an exact analytical solution was obtained in the limit of high thermal conductivity of the droplet (or up to the droplet size of -100 Å). This analytical solution incorporates parameters that are determined using in silico. experiments on isothermal droplet evaporation. We show that the kinetics of adiabatic evaporation and cooling of the droplet scales with the droplet size. Our estimates of the water-protemi droplet. freezing rate in the adiabatic regime in a vacuum chamber show that additional techniques for stabilizing the temperature inside the droplet should be used in order to study the conformational transitions of the protein in single molecules. Isothermal and quasi-isothermal conditions are most suitable for studying the conformational transitions upon object functioning. However, in this case it is necessary to take into account the effects of dehydration and rapid increase of ionic strength in an

Operating experience with the DRAGON High Temperature Reactor experiment

International Nuclear Information System (INIS)

Simon, R.A.; Capp, P.D.

2002-01-01

The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor

International Nuclear Information System (INIS)

Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes

2011-01-01

This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

Modelling of water sump evaporation in a CFD code for nuclear containment studies

Energy Technology Data Exchange (ETDEWEB)

Malet, J., E-mail: jeanne.malet@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Bessiron, M., E-mail: matthieu.bessiron@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Perrotin, C., E-mail: christophe.perrotin@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France)

2011-05-15

Highlights: We model sump evaporation in the reactor containment for CFD codes. The sump is modelled by an interface temperature and an evaporation mass flow-rate. These two variables are modelled using energy and mass balance. Results are compared with specific experiments in a 7 m3 vessel (Tonus Qualification ANalytique, TOSQAN). A good agreement is observed, for pressure, temperatures, mass flow-rates. - Abstract: During the course of a hypothetical severe accident in a pressurized water reactor (PWR), water can be collected in the sump containment through steam condensation on walls and spray systems activation. This water is generally under evaporation conditions. The objective of this paper is twofold: to present a sump model developed using external user-defined functions for the TONUS-CFD code and to perform a first detailed comparison of the model results with experimental data. The sump model proposed here is based on energy and mass balance and leads to a good agreement between the numerical and the experimental results. Such a model can be rather easily added to any CFD code for which boundary conditions, such as injection temperature and mass flow-rate, can be modified by external user-defined functions, depending on the atmosphere conditions.

Evaporation of lead and lithium from molten Pb-17Li - transport of aerosols

International Nuclear Information System (INIS)

Feuerstein, H.; Graebner, H.; Oschinski, J.; Horn, S.; Bender, S.

1991-01-01

Evaporation of Pb and Li from molten Pb-17Li was investigated between 350 and 800deg C in vacuum, argon and helium covergas. Results were also obtained from other experimental facilities. Similarities were found to observations from sodium cooled reactors. The results show that Pb and Li evaporate independent on each other. The two elements show different behavior along the transport pathway. Deposits of the evaporated metals contained between 0.2 and 98 at% Li. As in the reactor RAPSODIE for sodium, evaporation rates for lithium were smaller in helium than in argon, however evaporation rates of lead were the same in both gases. No aerosol problems will exist with normal blanket operation. Under experimental conditions, aerosol concentrations were in the range of 10 -9 to 10 -6 g/m 3 . Aerosols can easily be trapped with sintered metal filters. (orig.)

IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

International Nuclear Information System (INIS)

Sartori, Enrico; Blair Briggs, J.

2008-01-01

1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

Secondary decay of espalation in ADS reactors

International Nuclear Information System (INIS)

Rodrigues, Marcos Guedes; Santiago, A.J.; Silva, C.E. da

2013-01-01

We study the problem of evaporation in the context of nuclear spallation reactions in nuclear reactors ADS. The calculation was developed based on the theory of Weisskopf evaporation and in the model of thermal liquid drop. Evaporation affects the 'economy' of neutrons and the design of a ADS reactor in various aspects. It offers abundant amount of neutrons in the nuclear medium, with a wide energy range. For an excitation energy of 3 MeV/n a typical core evaporates about 10% of its mass in the form of light particles (mostly neutrons)

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-10-15

At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

Lessons from early experience in reactor development

International Nuclear Information System (INIS)

Allen, W.

1976-09-01

This paper deals with several issues in U.S. reactor development and demonstration experience. The focus is on the period between 1946 and 1963 during which the Atomic Energy Commission (AEC) guided early reactor research and development (R and D) and conducted the Power Reactor Demonstration Program

The experiences of research reactor accident to safety improvement

International Nuclear Information System (INIS)

Wiranto, S.

1999-01-01

The safety of reactor operation is the main factor in order that the nuclear technology development program can be held according the expected target. Several experience with research reactor incidents must be learned and understood by the nuclear program personnel, especially for operators and supervisors of RSG-GA. Siwabessy. From the incident experience of research reactor in the world, which mentioned in the book 'Experience with research reactor incidents' by IAEA, 1995, was concluded that the main cause of research reactor accidents is understandless about the safety culture by the nuclear installation personnel. With learn, understand and compare between this experiences and the condition of RSG GA Siwabessy is expended the operators and supervisors more attention about the safety culture, so that RSG GA Siwabessy can be operated successfull, safely according the expected target

Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

International Nuclear Information System (INIS)

1978-09-01

The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

Energy Technology Data Exchange (ETDEWEB)

Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

1999-08-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operation experience of the Indonesian multipurpose research reactor RSG-GAS

International Nuclear Information System (INIS)

Hastowo, Hudi; Tarigan, Alim

1999-01-01

RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

Operational and reliability experience with reactor instrumentation

International Nuclear Information System (INIS)

Dixon, F.; Gow, R.S.

1978-01-01

In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-08-31

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 ⁷LiF-BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

International Nuclear Information System (INIS)

Brown, Nicholas R.; Mueller, Donald E.; Patton, Bruce W.; Powers, Jeffrey J.

2016-01-01

Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 "7LiF-BeF_2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

The reactor antineutrino anomaly and low energy threshold neutrino experiments

Science.gov (United States)

Cañas, B. C.; Garcés, E. A.; Miranda, O. G.; Parada, A.

2018-01-01

Short distance reactor antineutrino experiments measure an antineutrino spectrum a few percent lower than expected from theoretical predictions. In this work we study the potential of low energy threshold reactor experiments in the context of a light sterile neutrino signal. We discuss the perspectives of the recently detected coherent elastic neutrino-nucleus scattering in future reactor antineutrino experiments. We find that the expectations to improve the current constraints on the mixing with sterile neutrinos are promising. We also analyze the measurements of antineutrino scattering off electrons from short distance reactor experiments. In this case, the statistics is not competitive with inverse beta decay experiments, although future experiments might play a role when compare it with the Gallium anomaly.

The detection, causes and repair of the small steam leaks in the PFR evaporator units

International Nuclear Information System (INIS)

Smedley, J.A.; Broomfield, A.M.; Anderson, R.

1984-01-01

The occurrence of a number of small steam leaks into the gas space above the sodium in the evaporator units of the UKAEA's Prototype Fast Reactor at Dounreay has had a significant impact on plant availability. The paper describes experience with the leak detection system and the phenomena which have caused the leaks and an outline is given of the measures which have been introduced to remedy the problem. (author)

Experiments and Modelling Techniques for Heat and Mass Transfer in Light Water Reactors

International Nuclear Information System (INIS)

Ambrosini, W.; Bucci, M.; Forgione, N.; Manfredini, A.; Oriolo, F.

2009-01-01

The paper summarizes the lesson learned from theoretical and experimental activities performed at the University of Pisa, Pisa, Italy, in past decades in order to develop a general methodology of analysis of heat and mass transfer phenomena of interest for nuclear reactor applications. An overview of previously published results is proposed, highlighting the rationale at the basis of the performed work and its relevant conclusions. Experimental data from different sources provided information for model development and assessment. They include condensation experiments performed at SIET (Piacenza, Italy) on the PANTHERS prototypical PCCS module, falling film evaporation tests for simulating AP600-like outer shell spraying conditions, performed at the University of Pisa, experimental data concerning condensation on finned tubes, collected by CISE (Piacenza, Italy) in the frame of the INCON EU Project, and experimental tests performed in the CONAN experimental facility installed at the University of Pisa. The experience gained in these activities is critically reviewed and discussed to highlight the relevant obtained conclusions and the perspectives for future work

Thermal denitrification of evaporators concentrates in reactor with fluidized bed

International Nuclear Information System (INIS)

Brugnot, C.

1993-11-01

As part of the treatments of liquid wastes coming from the Marcoule reprocessing plant, the study of a thermal denitrification process for evaporator concentrates has been chosen by the CEA/CEN Cadarache: the fluidized-bed calcination. This work presents the study of a calcination pilot-plant for wastes with a very high sodium nitrate content. After a reactional analysis carried out in a thermobalance on samples which are representative of the fluidized-bed compounds, the perfecting of many of the plant parameters - such as the solution injection system - was carried out on a scale-model at first. Then, it was verified on the pilot-plant, and some experiments have been carried out. A mathematical model for the particle growth inside the fluidized-bed is proposed. (author). 179 refs., 65 figs., 23 tabs

International Reactor Physics Experiment Evaluation (IRPhE) Project

International Nuclear Information System (INIS)

2013-01-01

The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

CANDU reactor experience: fuel performance

International Nuclear Information System (INIS)

Truant, P.T.; Hastings, I.J.

1985-07-01

Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

Experiences in stability testing of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Otaduy, P.J.

1986-01-01

The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

International Nuclear Information System (INIS)

Reuscher, J.A.; Pickard, P.S.

1976-01-01

The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

Experiment operations plan for the MT-4 experiment in the NRU reactor

International Nuclear Information System (INIS)

Russcher, G.E.; Wilson, C.L.; Parchen, L.J.; Marshall, R.K.; Hesson, G.M.; Webb, B.J.; Freshley, M.D.

1983-06-01

A series of thermal-hydraulic and cladding materials deformation experiments were conducted using light-water reactor fuel bundles as part of the Pacific Northwest Laboratory Loss-of-Coolant Accident (LOCA) Simulation Program. This report is the formal operations plan for MT-4 - the fourth materials deformation experiment conducted in the National Research Universal (NRU) reactor, Chalk River, Ontario, Canada. A major objective of MT-4 was to simulate a pressurized water reactor LOCA that could induce fuel rod cladding deformation and rupture due to a short-term adiabatic transient and a peak fuel cladding temperature of 1200K (1700 0 F)

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

Sodium evaporation into a forced argon flow

International Nuclear Information System (INIS)

Kumada, Toshiaki; Kasahara, Fumio; Ishiguro, Ryoji

1975-01-01

Evaporation from a rectangular sodium free surface into an argon flow was measured. Tests were carried out with varying sodium temperature, argon velocity and argon temperature respectively under conditions of fog formation being possible. In order to clarify the enhancement of evaporation by fog formation, convection heat transfer from a plate of the same geometry into an air flow was also measured. The evaporation rate and Sherwood number were compared with those predicted by both the heat transfer experiment and the theory proposed by Hill and Szekely, and also a comparison was run with the previously reported experimental results of sodium evaporation. As a result it was shown that the sodium evaporation rate in this experiment is at least four times as large as that predicted by the heat transfer experiment and varies almost linearly with the heat transfer rate and the sodium vapour pressure. (auth.)

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (II). The second experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-01-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the second reactor noise experiment using pseudo random binary sequences (PRBS) was performed on August 30, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 50% of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (III). The third experiment

International Nuclear Information System (INIS)

Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

1995-03-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the third reactor noise experiment using pseudo random binary sequences (PRBS) was performed on September 16, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 70% of reactor power and under a normal sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Tritium experience in the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Hogan, J.

1998-01-01

Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors

Laboratory experiments on solute transport in bimodal porous media under cyclic precipitation-evaporation boundary conditions

Science.gov (United States)

Cremer, Clemens; Neuweiler, Insa

2016-04-01

Flow and solute transport in the shallow subsurface is strongly governed by atmospheric boundary conditions. Erratically varying infiltration and evaporation cycles lead to alternating upward and downward flow, as well as spatially and temporally varying water contents and associated hydraulic conductivity of the prevailing materials. Thus presenting a highly complicated, dynamic system. Knowledge of subsurface solute transport processes is vital to assess e.g. the entry of, potentially hazardous, solutes to the groundwater and nutrient uptake by plant roots and can be gained in many ways. Besides field measurements and numerical simulations, physical laboratory experiments represent a way to establish process understanding and furthermore validate numerical schemes. With the aim to gain a better understanding and to quantify solute transport in the unsaturated shallow subsurface under natural precipitation conditions in heterogeneous media, we conduct physical laboratory experiments in a 22 cm x 8 cm x 1 cm flow cell that is filled with two types of sand and apply cyclic infiltration-evaporation phases at the soil surface. Pressure at the bottom of the domain is kept constant. Following recent studies (Lehmann and Or, 2009; Bechtold et al., 2011a), heterogeneity is introduced by a sharp vertical interface between coarse and fine sand. Fluorescent tracers are used to i) qualitatively visualize transport paths within the domain and ii) quantify solute leaching at the bottom of the domain. Temporal and spatial variations in water content during the experiment are derived from x-ray radiographic images. Monitored water contents between infiltration and evaporation considerably changed in the coarse sand while the fine sand remained saturated throughout the experiments. Lateral solute transport through the interface in both directions at different depths of the investigated soil columns were observed. This depended on the flow rate applied at the soil surface and

Review of irradiation experiments for water reactor safety research

International Nuclear Information System (INIS)

Tobioka, Toshiaki

1977-02-01

A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

Energy Technology Data Exchange (ETDEWEB)

Gutmann, H

1972-06-05

The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

STUDI EKSPERIMENTAL FALLING FILM EVAPORATOR PADA EVAPORASI NIRA KENTAL

Directory of Open Access Journals (Sweden)

Medya Ayunda Fitri

2016-06-01

Full Text Available Falling film evaporator is a constructed equipment for concentrating dilute solution that are sensitive to heat flowing form a thin film. This research aims to study the evaporation of cane juice concentrated with air flow on falling film evaporator and knowing evaporation rate occured in falling film evaporator used. In the process, cane juice from plant pumped to the falling film evaporator that used in this experiment. This research used concentrated cane juice and air flow rate for variables of this experiment. Cane juice flow from top of evaporator through distributor to form thin film and air flow from the bottom of evaporator. After that, temperatur of pipe wall, inlet and outlet temperature of cane juice and air were measured. This experiment concluded that the highest concentration of outlet solution is 59 brix for liquid flow rate 154 l/h and air flow rate 10 m3/h, and the other hand inlet solution concentration 51 brix. Optimum evaporation rate is 35 kg/m2.h for 51 brix and air flow rate 10 m3/h.

Treatment of liquid radioactive waste: Evaporation

International Nuclear Information System (INIS)

Pfeiffer, R.

1982-01-01

About 10.000 m 3 of low active liquid waste (LLW) arise in the Nuclear Research Center Karlsruhe. Chemical contents of this liquid waste are generally not declared. Resulting from experiments carried out in the Center during the early sixties, the evaporator facility was built in 1968 for decontamination of LLW. The evaporators use vapor compression and concentrate recirculation in the evaporator sump by pumps. Since 1971 the medium active liquid waste (MLW) from the Karlsruhe Reprocessing Plant (WAK) was decontaminated in this evaporator facility, too. By this time the amount of low liquid waste (LLW) had been decontaminated without mentionable interruptions. Afterwards a lot of interruptions of operations occurred, mainly due to leakages of pumps, valves and pipes. There was also a very high radiation level for the operating personnel. As a consequence of this experience a new evaporator facility for decontamination of medium active liquid waste was built in 1974. This facility started operation in 1976. The evaporator has natural circulation and is heated by steam through a heat exchanger. (orig./RW)

International Experience with Fast Reactor Operation & Testing

International Nuclear Information System (INIS)

Sackett, John I.; Grandy, C.

2013-01-01

Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

International Nuclear Information System (INIS)

Beech, D.J.; May, R.

2000-01-01

In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

Use of the transpiration method to study polonium evaporation from liquid lead-bismuth eutectic at high temperature

International Nuclear Information System (INIS)

Prieto, Borja Gonzalez; Lim, Jun; Rosseel, Kris; Bosch, Joris van den; Aerts, Alexander; Martens, Johan; Rizzi, Matthias; Neuhausen, Joerg

2014-01-01

Qualitative and quantitative understanding of Po volatilization under different conditions is of key importance for safety assessments of lead-bismuth eutectic (LBE) based nuclear reactors, spallation targets and accelerator driven systems. In this work we explore the possibilities of the transpiration method in combination with simple models to study the equilibrium and kinetics of Po evaporation from highly diluted solutions in lead-bismuth eutectic between 600 and 1000 C in Ar/5% H 2 and Ar. On the basis of evaporation experiments at various carrier gas flow rates, we identified the conditions of vapor saturation allowing the determination of equilibrium constants. From the limiting behavior at high flow rates, values for the maximal evaporation rate of Po from LBE were estimated. Measurements of evaporation as a function of time were consistent with the assumption that polonium dissolved in LBE obeys Henry's law. A theoretical analysis furthermore suggested that diffusion of polonium in LBE was not a rate limiting factor for evaporation under vapor saturation conditions. Newly determined values for the Henry constant of Po in LBE between 600 and 1000 C were consistent with previously derived correlations.

Advanced test reactor testing experience-past, present and future

International Nuclear Information System (INIS)

Marshall, Frances M.

2006-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

Channel type reactors with supercritical water coolant. Russian experience

International Nuclear Information System (INIS)

Kuznetsov, Y.N.; Gabaraev, B.A.

2003-01-01

Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

Dynamics of Soil Water Evaporation during Soil Drying in the Presence of a Shallow Water Table: Laboratory Experiment and Numerical Analysis

Science.gov (United States)

Han, J.; Lin, J.; Liu, P.; Li, W.

2017-12-01

Evaporation from a porous medium plays a key role in hydrological, agricultural, environmental, and engineering applications. Laboratory and numerical experiments were conducted to investigate the evolution of soil water evaporation during a continuous drying event. Simulated soil water contents and temperatures by the calibrated model well reproduced measured values at different depths. Results show that the evaporative drying process could be divided into three stages, beginning with a relatively high evaporation rate during stage 1, followed by a lower rate during transient stage and stage 2, and finally maintaining a very low and constant rate during stage 3. The condensation zone was located immediately below the evaporation zone in the profile. Both peaks of evaporation and condensation rate increased rapidly during stage 1 and transition stage, decreased during stage 2, and maintained constant during stage 3. The width of evaporation zone kept a continuous increase during stages 1 and 2 and maintained a nearly constant value of 0.68 cm during stage 3. When the evaporation zone totally moved into the subsurface, a dry surface layer (DSL) formed above the evaporation zone at the end of stage 2. The width of DSL also presented a continuous increase during stage 2 and kept a constant value of 0.71 cm during stage 3. Although the magnitude of condensation zone was much smaller than that for the evaporation zone, the importance of the contribution of condensation zone to soil water dynamics should not be underestimated. Results from our experiment and numerical simulation show that this condensation process resulted in an unexpected and apparent water content increase in the middle of vadose zone profile.

Detecting Dark Photons with Reactor Neutrino Experiments

Science.gov (United States)

Park, H. K.

2017-08-01

We propose to search for light U (1 ) dark photons, A', produced via kinetically mixing with ordinary photons via the Compton-like process, γ e-→A'e-, in a nuclear reactor and detected by their interactions with the material in the active volumes of reactor neutrino experiments. We derive 95% confidence-level upper limits on ɛ , the A'-γ mixing parameter, ɛ , for dark-photon masses below 1 MeV of ɛ reactors as potential sources of intense fluxes of low-mass dark photons.

Validation of the RALOC-mod.4 thermal-hydraulics code on evaporation transients in the Phebus containment

International Nuclear Information System (INIS)

Spitz, P.B.; Lemoine, F.; Tirini, S.

1997-01-01

IPSN (Nuclear Protection and Safety Institute) and GRS (Gesellschaft fur Anlagen und Reaktorsicherheit Schwertnergasse 1) are developing the ESCADRE-ASTEC systems of codes devoted to the prediction of the behaviour of water-cooled reactors during a severe accident. The RALOC-mod 4 code belongs to this system and is specifically devoted to containment thermal-hydraulics studies. IPSN has designed a Thermal Hydraulic Containment Test Program in support to the Phebus Fission Product Test Program/2/. Evaporation tests have been recently performed in the Phebus containment test facility. The objective of this work is to assess against these tests the capability of the RALOC -mod 4 code to capture the phenomena observed in these experiments and more particularly the evaporation heat transfer and wall heat transfers. (DM)

Storage experience in Hungary with fuel from research reactors

International Nuclear Information System (INIS)

Gado, J.; Hargitai, T.

1996-01-01

In Hungary several critical assemblies, a training reactor and a research reactor have been in operation. The fuel used in the research and training reactors are of Soviet origin. Though spent fuel storage experience is fairly good, medium and long term storage solutions are needed. (author)

Investigating the spectral anomaly with different reactor antineutrino experiments

Directory of Open Access Journals (Sweden)

C. Buck

2017-02-01

Full Text Available The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in U235 with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted U235 spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

Operational experience of the Marcoule reactors

International Nuclear Information System (INIS)

Conte, F.

1963-01-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

Evaporative cooling in polymer electrolyte fuel cell

Energy Technology Data Exchange (ETDEWEB)

Shimotori, S; Sonai, A [Toshiba Corp. Tokyo (Japan)

1996-06-05

The concept of the evaporative cooling for the internally humidified PEFC was confirmed by the experiment. The evaporative cooling rates at the anode and the cathode were mastered under the various temperatures and air utilizations. At a high temperature the proportion of the evaporative cooling rate to the heat generation rate got higher, the possibility of the evaporative cooling was demonstrated. 2 refs., 7 figs., 1 tab.

Leak before break experience in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Moan, G.D.; Coleman, C.E.

1988-04-01

The paper describes how the requirements for Leak-Before-Break are met in CANDU reactors. The requirements are based on operational and laboratory experience. After the onset of leakage in a fuel channel from a delayed hydride crack, time is available to the operator to take action before the crack grows to an unstable length. The time available is calculated using different models which use crack growth data from small specimen tests. When the results from crack growth behaviour experiments, carried out on components removed from reactor are used in the model, the time available for operator response is about 100 hours

Fast breeder reactors: can we learn from experience

International Nuclear Information System (INIS)

Keck, O.

1981-01-01

An economic analysis of FBRs, in particular the long-term benefits to be expected, with reference to the experience of the West German fast breeder reactor programme suggests ways of bringing more realism into governmental decisions on the development of new reactor types. It is suggested that if reactor manufacturers and utilities financed commercial-size demonstration plants from their own funds, then the government would get more realistic advice. (U.K.)

Coupling of AST-500 heating reactors with desalination facilities

International Nuclear Information System (INIS)

Kourachenkov, A.V.

1998-01-01

The general issues regarding NHR and desalination facility joint operation for potable water production are briefly considered. AST-500 reactor plant and DOU GTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. Similarity of NHR operation for a heating grid and a desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author)

Solubility of plutonium and waste evaporation

International Nuclear Information System (INIS)

Karraker, D.G.

1993-01-01

Chemical processing of irradiated reactor elements at the Savannah River Site separates uranium, plutonium and fission products; fission products and process-added chemicals are mixed with an excess of NaOH and discharged as a basic slurry into large underground tanks for temporary storage. The slurry is composed of base-insoluble solids that settle to the bottom of the tank; the liquid supemate contains a mixture of base-soluble chemicals--nitrates, nitrites aluminate, sulfate, etc. To conserve space in the waste tanks, the supemate is concentrated by evaporation. As the evaporation proceeds, the solubilities of some components are exceeded, and these species crystallize from solution. Normally, these components are soluble in the hot solution discharged from the waste tank evaporator and do not crystallize until the solution cools. However, concern was aroused at West Valley over the possibility that plutonium would precipitate and accumulate in the evaporator, conceivably to the point that a nuclear accident was possible. There is also a concern at SRS from evaporation of sludge washes, which arise from washing the base-insoluble solids (open-quote sludge close-quote) with ca. 1M NaOH to reduce the Al and S0 4 -2 content. The sludge washes of necessity extract a low level of Pu from the sludge and are evaporated to reduce their volume, presenting the possibility of precipitating Pu. Measurements of the solubility of Pu in synthetic solutions of similar composition to waste supernate and sludge washes are described in this report

Licensing experience of the HTR-10 test reactor

International Nuclear Information System (INIS)

Sun, Y.; Xu, Y.

1996-01-01

A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

Review of experiments for research reactors - approved 1974

International Nuclear Information System (INIS)

Anon.

1977-01-01

This standard establishes guidelines for the review and approval of experiments performed at research reactor facilities. This standard identifies the major areas that shall be reviewed for each experiment to ensure that it (a) falls within the limits delineated in the technical specifications, (b) does not present an unreviewed safety question as defined in 10 CFR Section 50.59 π2-, (c) does not constitute a threat to the health and safety of any individual or group of individuals, and (d) does not constitute a hazard to the reactor facility or other equipment. In addition, this standard recommends a system for classifying experiments to establish levels of review and approval commensurate with the level of risk inherent in the experiment

Evaporation from bare ground with different water-table depths based on an in-situ experiment in Ordos Plateau, China

Science.gov (United States)

Zhang, Zaiyong; Wang, Wenke; Wang, Zhoufeng; Chen, Li; Gong, Chengcheng

2018-03-01

The dynamic processes of ground evaporation are complex and are related to a multitude of factors such as meteorological influences, water-table depth, and materials in the unsaturated zone. To investigate ground evaporation from a homogeneous unsaturated zone, an in-situ experiment was conducted in Ordos Plateau of China. Two water-table depths were chosen to explore the water movement in the unsaturated zone and ground evaporation. Based on the experimental and calculated results, it was revealed that (1) bare ground evaporation is an atmospheric-limited stage for the case of water-table depth being close to the capillary height; (2) the bare ground evaporation is a water-storage-limited stage for the case of water-table depth being beyond the capillary height; (3) groundwater has little effect on ground-surface evaporation when the water depth is larger than the capillary height; and (4) ground evaporation is greater at nighttime than that during the daytime; and (5) a liquid-vapor interaction zone at nearly 20 cm depth is found, in which there exists a downward vapor flux on sunny days, leading to an increasing trend of soil moisture between 09:00 to 17:00; the maximum value is reached at midday. The results of this investigation are useful to further understand the dynamic processes of ground evaporation in arid areas.

Reactor wall in thermonuclear device

International Nuclear Information System (INIS)

Shibui, Masanao.

1988-01-01

Purpose: To always monitor the life of armours in reactor walls and automatically shutdown the reactor if it should be operated in excess of the limit of use. Constitution: Monitoring material of lower melting point than armours (for example beryllium pellets) as one of the reactor wall constituents of a thermonuclear device are embedded in a region leaving the thickness corresponding to the allowable abrasion of the armour. In this structure, if the armours are abrased due to particle loads of a plasma and the abrasion exceeds a predetermined allowable level, the monitoring material is exposed to the plasma and melted and evaporated. Since this can be detected by impurity monitors disposed in the reactor, it is possible to recognize the limit for the working life of the armours. If the thermonuclear reactor should be operated accidentally exceeding the life of the armours, since a great amount of the monitoring materials have been evaporated, they flow into the plasma to increase the plasma radiation loss thereby automatically eliminate the plasma. (K.M.)

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

International Nuclear Information System (INIS)

Bess, John D.; Gullifor, Jim

2015-03-01

The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

Experience and prospects for developing research reactors of different types

International Nuclear Information System (INIS)

Kuatbekov, R.P.; Tretyakov, I.T.; Romanov, N.V.; Lukasevich, I.B.

2015-01-01

NIKIET has a 60-year experience in the development of research reactors. Altogether, there have been more than 25 NIKIET-designed plants of different types built in Russia and 20 more in other countries, including pool-type water-cooled and water moderated research reactors, tank-type and pressure-tube research reactors, pressurized high-flux, heavy-water, pulsed and other research reactors. Most of the research reactors were designed as multipurpose plants for operation at research centers in a broad range of applications. Besides, unique research reactors were developed for specific application fields. Apart from the experience in the development of research reactor designs and the participation in the reactor construction, a unique amount of knowledge has been gained on the operation of research reactors. This makes it possible to use highly reliable technical solutions in the designs of new research reactors to ensure increased safety, greater economic efficiency and maintainability of the reactor systems. A multipurpose pool-type research reactor of a new generation is planned to be built at the Center for Nuclear Energy Science & Technology (CNEST) in the Socialist Republic of Vietnam to be used to support a spectrum of research activities, training of skilled personnel for Vietnam nuclear industry and efficient production of isotopes. It is exactly the applications a research reactor is designed for that defines the reactor type, design and capacity, and the selection of fuel and components subject to all requirements of industry regulations. The design of the new research reactor has a great potential in terms of upgrading and installation of extra experimental devices. (author)

Technical report on operating experience with boiling water reactor offgas systems

International Nuclear Information System (INIS)

Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

1978-03-01

Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-01

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Song, C. H.; Chung, M. K.; Park, C. K. and others

2005-04-15

The objectives of the project are to study thermal hydraulic characteristics of reactor primary system for the verification of the reactor safety and to evaluate new safety concepts of new safety design features. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. Followings are main research topics; - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation Load and Thermal Mixing in the IRWST - Development of Thermal-Hydraulic Models for Two-Phase Flow - Development of Measurement Techniques for Two-Phase Flow - Supercritical Reactor T/H Characteristics Analysis From the above experimental and analytical studies, new safety design features of the advanced power reactors were verified and lots of the safety issues were also resolved.

Operating experience of the EBR-II steam generating system

International Nuclear Information System (INIS)

Buschman, H.W.; Penney, W.H.; Quilici, M.D.; Radtke, W.H.

1981-01-01

The Experimental Breeder Reactor II (EBR-II) is a Liquid Metal Fast Breeder Reactor (LMFBR) with integrated power producing capability. Superheated steam is produced by eight natural circulation evaporators, two superheaters, and a conventional steam drum. Steam throttle conditions are 438 C (820 F) and 8.62 MPa (1250 psi). The designs of the evaporators and superheaters are essentially identical; both are counterflow units with low pressure nonradioactive sodium on the shell side. Safety and reliability are maximized by using duplex tubes and tubesheets. The performance of the system has been excellent and essentially trouble free. The operating experience of EBR-II provides confidence that the technology can be applied to commercial LMFBR's for an abundant supply of energy for the future. 5 refs

The continuous similarity model of bulk soil-water evaporation

Science.gov (United States)

Clapp, R. B.

1983-01-01

The continuous similarity model of evaporation is described. In it, evaporation is conceptualized as a two stage process. For an initially moist soil, evaporation is first climate limited, but later it becomes soil limited. During the latter stage, the evaporation rate is termed evaporability, and mathematically it is inversely proportional to the evaporation deficit. A functional approximation of the moisture distribution within the soil column is also included in the model. The model was tested using data from four experiments conducted near Phoenix, Arizona; and there was excellent agreement between the simulated and observed evaporation. The model also predicted the time of transition to the soil limited stage reasonably well. For one of the experiments, a third stage of evaporation, when vapor diffusion predominates, was observed. The occurrence of this stage was related to the decrease in moisture at the surface of the soil. The continuous similarity model does not account for vapor flow. The results show that climate, through the potential evaporation rate, has a strong influence on the time of transition to the soil limited stage. After this transition, however, bulk evaporation is independent of climate until the effects of vapor flow within the soil predominate.

Experience in operation of heavy water reactors

International Nuclear Information System (INIS)

Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

1999-01-01

The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

Experience in utilizing research reactors in Yugoslavia

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J.; Raisic, N. [Boris Kidric Institute of Nuclear Sciences VINCA, Belgrade (Yugoslavia); Copic, M.; Gabrovsek, Z. [Jozef Stefan Institute Ljubljana (Yugoslavia)

1972-07-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Experience in utilizing research reactors in Yugoslavia

International Nuclear Information System (INIS)

Pop-Jordanov, J.; Raisic, N.; Copic, M.; Gabrovsek, Z.

1972-01-01

The nuclear institutes in Yugoslavia possess three research reactors. Since 1958, two heavy-water reactors have been in operation at the 'Boris Kidric' Institute, a zero-power reactor RB and a 6. 5-MW reactor RA. At the Jozef Stefan Institute, a 250-kW TRIGA Mark II reactor has been operating since 1966. All reactors are equipped with the necessary experimental facilities. The main activities based on these reactors are: (1) fundamental research in solid-state and nuclear physics; (2) R and D activities related to nuclear power program; and (3) radioisotope production. In fundamental physics, inelastic neutron scattering and diffraction phenomena are studied by means of the neutron beam tubes and applied to investigations of the structures of solids and liquids. Valuable results are also obtained in n - γ reaction studies. Experiments connected with the fuel -element development program, owing to the characteristics of the existing reactors, are limited to determination of the fuel element parameters, to studies on the purity of uranium, and to a small number of capsule irradiations. All three reactors are also used for the verification of different methods applied in the analysis of power reactors, particularly concerning neutron flux distributions, the optimization of reactor core configurations and the shielding effects. An appreciable irradiation space in the reactors is reserved for isotope production. Fruitful international co-operation has been established in all these activities, on the basis of either bilateral or multilateral arrangements. The paper gives a critical analysis of the utilization of research reactors in a developing country such as Yugoslavia. The investments in and the operational costs of research reactors are compared with the benefits obtained in different areas of reactor application. The impact on the general scientific, technological and educational level in the country is also considered. In particular, an attempt is made ro

Evaporation of nanofluid droplet on heated surface

Directory of Open Access Journals (Sweden)

Yeung Chan Kim

2015-04-01

Full Text Available In this study, an experiment on the evaporation of nanofluid sessile droplet on a heated surface was conducted. A nanofluid of 0.5% volumetric concentration mixed with 80-nm-sized CuO powder and pure water were used for experiment. Droplet was applied to the heated surface, and images of the evaporation process were obtained. The recorded images were analyzed to find the volume, diameter, and contact angle of the droplet. In addition, the evaporative heat transfer coefficient was calculated from experimental result. The results of this study are summarized as follows: the base diameter of the droplet was maintained stably during the evaporation. The measured temperature of the droplet was increased rapidly for a very short time, then maintained constantly. The nanofluid droplet was evaporated faster than the pure water droplet under the experimental conditions of the same initial volume and temperature, and the average evaporative heat transfer coefficient of the nanofluid droplet was higher than that of pure water. We can consider the effects of the initial contact angle and thermal conductivity of nanofluid as the reason for this experimental result. However, the effect of surface roughness on the evaporative heat transfer of nanofluid droplet appeared unclear.

Coupling of AST-500 heating reactors with desalination facilities

International Nuclear Information System (INIS)

Gureyeva, L.V.; Egorov, V.V.; Podberezniy, V.L.

1997-01-01

The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab

Coupling of AST-500 heating reactors with desalination facilities

Energy Technology Data Exchange (ETDEWEB)

Gureyeva, L V; Egorov, V V [OKBM, Nizhny Novgorod (Russian Federation); Podberezniy, V L [Scientific Research Inst. of Machine Building, Ekaterinburg (Russian Federation)

1997-09-01

The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab.

Development of the design of the High Temperature Gas Cooled Reactor experiment

International Nuclear Information System (INIS)

Lockett, G.E.; Huddle, R.A.U.

1960-01-01

Early in 1956 a small team was formed at the Atomic Energy Research Establishment, Harwell, to investigate the possibilities of the High Temperature Gas Cooled (H.T.G.C.) Reactor System. Although the primary objective of this team was to carry out a feasibility study of the system as a whole, it soon became apparent that, in addition to design studies and economic surveys of power producing reactors, the most appropriate approach to such a novel system was to carry out a design study of a relatively small (10 to 20 M.W.) Reactor Experiment, together with the necessary research and development work associated with such a reactor. This work proceeded within the U.K.A.E.A. during the three following years, and it was felt that realistic design proposals could be put forward with sufficient confidence to justify the detailed design and construction of a 20 M.W. Reactor Experiment. In April 1959 responsibility for this Reactor Experiment was taken over by the O.E.E.C. High Temperature Gas Cooled Reactor Project, the DRAGON Project, at the Atomic Energy Establishment, Winfrith, Dorset. In this Paper the research, development, and design work is reviewed, and the proposals for the Reactor Experiment are summarised. (author)

Improvements of evaporation drag model

International Nuclear Information System (INIS)

Li Xiaoyan; Yang Yanhua; Xu Jijun

2004-01-01

A special observable experiment facility has been established, and a series of experiments have been carried out on this facility by pouring one or several high-temperature particles into a water pool. The experiment has verified the evaporation drag model, which believe the non-symmetric profile of the local evaporation rate and the local density of the vapor would bring about a resultant force on the hot particle so as to resist its motion. However, in Yang's evaporation drag model, radiation heat transfer is taken as the only way to transfer heat from hot particle to the vapor-liquid interface and all of the radiation energy is deposited on the vapor-liquid interface, thus contributing to the vaporization rate and mass balance of the vapor film. So, the heat conduction and the heat convection are taken into account in improved model. At the same time, the improved model given by this paper presented calculations of the effect of hot particles temperature on the radiation absorption behavior of water

Evaporative removal of sodium: interim progress report and preliminary facility specification

International Nuclear Information System (INIS)

Welch, F.H.

1978-01-01

A summary of the current Evaporative Removal of Sodium (ERNA) activities at the Energy Systems Group is presented. Also included is a review of earlier work on sodium evaporation. As a result of this work it was concluded that the ERNA process was extremely successful and worthy of future consideration as a recognized process for reactor components. Also included in the report is a Preliminary Outline Specification for a large facility to remove sodium from full size CRBR fuel rod assemblies

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

Grover, S. Blaine

2009-01-01

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

Non-standard interaction effects at reactor neutrino experiments

International Nuclear Information System (INIS)

Ohlsson, Tommy; Zhang, He

2009-01-01

We study non-standard interactions (NSIs) at reactor neutrino experiments, and in particular, the mimicking effects on θ 13 . We present generic formulas for oscillation probabilities including NSIs from sources and detectors. Instructive mappings between the fundamental leptonic mixing parameters and the effective leptonic mixing parameters are established. In addition, NSI corrections to the mixing angles θ 13 and θ 12 are discussed in detailed. Finally, we show that, even for a vanishing θ 13 , an oscillation phenomenon may still be observed in future short baseline reactor neutrino experiments, such as Double Chooz and Daya Bay, due to the existences of NSIs

Brookhaven Reactor Experiment Control Facility, a distributed function computer network

International Nuclear Information System (INIS)

Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

1975-11-01

A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

Evaporation

International Nuclear Information System (INIS)

Delaney, B.T.; Turner, R.J.

1989-01-01

Evaporation has long been used as a unit operation in the manufacture of various products in the chemical-process industries. In addition, it is currently being used for the treatment of hazardous wastes such as radioactive liquids and sludges, metal-plating wastes, and other organic and inorganic wastes. Design choice is dependent on the liquid to be evaporated. The three most common types of evaporation equipment are the rising-film, falling-film, and forced-circulation evaporators. The first two rely on boiling heat transfer and the latter relies on flash vaporization. Heat exchangers, flash tanks, and ejectors are common auxiliary equipment items incorporated with evaporator bodies to complete an evaporator system. Properties of the liquid to be evaporated are critical in final selection of an appropriate evaporator system. Since operating costs are a significant factor in overall cost, heat-transfer characteristics and energy requirements are important considerations. Properties of liquids which are critical to the determination of final design include: heat capacity, heat of vaporization, density, thermal conductivity, boiling point rise, and heat-transfer coefficient. Evaporation is an expensive technology, both in terms of capital costs and operating costs. Additionally, mechanical evaporation produces a condensate and a bottoms stream, one or both of which may require further processing or disposal. 3 figs

Drop evaporation and triple line dynamics

Science.gov (United States)

Sobac, Benjamin; Brutin, David; Gavillet, Jerome; Université de Provence Team; Cea Liten Team

2011-03-01

Sessile drop evaporation is a phenomenon commonly came across in nature or in industry with cooling, paintings or DNA mapping. However, the evaporation of a drop deposited on a substrate is not completely understood due to the complexity of the problem. Here we investigate, with several nano-coating of the substrate (PTFE, SiOx, SiOc and CF), the influence of the dynamic of the triple line on the evaporation process. The experiment consists in analyzing simultaneously the motion of the triple line, the kinetics of evaporation, the internal thermal motion and the heat and mass transfer. Measurements of temperature, heat-flux and visualizations with visible and infrared cameras are performed. The dynamics of the evaporative heat flux appears clearly different depending of the motion of the triple line

New safety experiments in decommissioned superheated steam reactor at Karlstein

International Nuclear Information System (INIS)

Koerting, K.

1986-01-01

This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

International Nuclear Information System (INIS)

Agnolet, G.; Baker, W.; Barker, D.; Beck, R.; Carroll, T.J.; Cesar, J.; Cushman, P.; Dent, J.B.; De Rijck, S.; Dutta, B.; Flanagan, W.; Fritts, M.; Gao, Y.; Harris, H.R.; Hays, C.C.; Iyer, V.

2017-01-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5–20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

Background studies for the MINER Coherent Neutrino Scattering reactor experiment

Energy Technology Data Exchange (ETDEWEB)

Agnolet, G.; Baker, W. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Barker, D. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Beck, R. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Carroll, T.J.; Cesar, J. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Cushman, P. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Dent, J.B. [Department of Physics, University of Louisiana at Lafayette, Lafayette, LA 70504 (United States); De Rijck, S. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Dutta, B. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Flanagan, W. [Department of Physics, University of Texas at Austin, Austin, TX 78712 (United States); Fritts, M. [School of Physics & Astronomy, University of Minnesota, Minneapolis, MN 55455 (United States); Gao, Y. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Department of Physics & Astronomy, Wayne State University, Detroit 48201 (United States); Harris, H.R.; Hays, C.C. [Department of Physics and Astronomy, and the Mitchell Institute for Fundamental Physics and Astronomy, Texas A& M University, College Station, TX 77843 (United States); Iyer, V. [School of Physical Sciences, National Institute of Science Education and Research, Jatni - 752050 (India); and others

2017-05-01

The proposed Mitchell Institute Neutrino Experiment at Reactor (MINER) experiment at the Nuclear Science Center at Texas A&M University will search for coherent elastic neutrino-nucleus scattering within close proximity (about 2 m) of a 1 MW TRIGA nuclear reactor core using low threshold, cryogenic germanium and silicon detectors. Given the Standard Model cross section of the scattering process and the proposed experimental proximity to the reactor, as many as 5–20 events/kg/day are expected. We discuss the status of preliminary measurements to characterize the main backgrounds for the proposed experiment. Both in situ measurements at the experimental site and simulations using the MCNP and GEANT4 codes are described. A strategy for monitoring backgrounds during data taking is briefly discussed.

Change of neutron flow sensors effectiveness in the course of reactor experiments

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Kotov, V.M.; Zhotabaev, Zh.R.

2007-01-01

Full text: IGR reactor is a reactor of thermal capacity type. During the operation, uranium-graphite core can be heated up to 1500 deg. C and reactivity can be changed considerably. Core dimensions are comparatively small. Amount of control rods, providing required reactivity, is not big as well. Increasing of core temperature leads to the rise of neutrons path length in its basic material - graphite. Change of temperature is not even. All this causes the non-conservation of neutron flows ratio in irradiated sample and in the place of reactor power sensors installation. Deviations in this ratio were registered during the number of reactor experiments. Empiric corrections can be introduced in order to decrease influence of change of neutron flow effectiveness upon provision of required parameters of investigated matters load. However, dependence of these corrections upon many factors can lead to the increasing of instability of process control. Previous experiment-calculated experiments showed inequality of neutron field in the place of sensors location (up to tens of percent), low effectiveness of experimental works, carried out without access to the individual reactor laying elements. Imperfection during the experiment was an idea of possibility to connect distribution of out of reactor neutron flow and control rods position. Subsequent analysis showed that for the development of representative phenomenon model it is necessary to take into account reactor operation dynamic subject to unevenness of heating of individual laying parts. Elemental calculations showed that temperature laying effects in the change of neutron outer field are great. Algorithm of calculations for the change of outer filed and field of investigated fabrication includes calculation of neutron-physic reactor characteristics interlacing with calculations of thermal-physic reactor characteristics, providing correlation of temperature fields for neutron-physic calculations. In the course of such

DOE's foreign research reactor transportation services contract: Perspective and experience

International Nuclear Information System (INIS)

Patterson, John

1997-01-01

DOE committed to low- and moderate-income countries participating in the foreign research reactor spent fuel returns program that the United States government would provide for the transportation of the spent fuel. In fulfillment of that commitment, DOE entered into transportation services contracts with qualified, private-sector firms. NAC will discuss its experience as a transportation services provider, including range of services available to the foreign reactors, advantages to DOE and to the foreign research reactors, access to contract services by high income countries and potential advantages, and experience with initial tasks performed under the contract. (author)

Calculation to experiment comparison of SPND signals in various nuclear reactor environments

Energy Technology Data Exchange (ETDEWEB)

Barbot, Loic; Radulovic, Vladimir; Fourmentel, Damien [CEA, DEN, DER, Instrumentation, Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance, (France); Snoj, Luka [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana, (Slovenia); Tarchalski, Mikolaj [National Centre for Nuclear Research, ulica Andrzeja Soltana 7, 05-400 Otwock (Swierk), (Poland); Dewynter-Marty, Veronique [CEA, DEN, DANS, DRSN, SIREN, LESCI, Saclay, F-91191 Gif sur Yvette, (France); Malouch, Fadhel [CEA, DEN, DANS, DM2S, SERMA, Saclay, F-91191 Gif sur Yvette, (France)

2015-07-01

In the perspective of irradiation experiments in the future Jules Horowitz Reactor (JHR), the Instrumentation Sensors and Dosimetry Laboratory of CEA Cadarache (France) is developing a numerical tool for SPND design, simulation and operation. In the frame of the SPND numerical tool qualification, dedicated experiments have been performed both in the Slovenian TRIGA Mark II reactor (JSI) and very recently in the French CEA Saclay OSIRIS reactor, as well as a test of two detectors in the core of the Polish MARIA reactor (NCBJ). A full description of experimental set-ups and neutron-gamma calculations schemes are provided in the first part of the paper. Calculation to experiment comparison of the various SPNDs in the different reactors is thoroughly described and discussed in the second part. Presented comparisons show promising final results. (authors)

Present and future oscillation experiments at reactors

International Nuclear Information System (INIS)

Mikaehlyan, L.A.

2001-01-01

A report is presented on recent progress and developments (since the NANP'99 Conference) in the current and future long baseline (∼100 - 800 km) oscillation experiments at reactors. These experiments, under certain assumptions, can fully reconstruct the internal mass structure of the electron neutrino and provide a laboratory test of solar and atmospheric neutrino problems

Liquid jet experiments: relevance to inertial confinement fusion reactors

International Nuclear Information System (INIS)

Hoffman, M.A.

1981-01-01

In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

Is evaporative colling important for shallow clouds?

Science.gov (United States)

Gentine, P.; Park, S. B.; Davini, P.; D'Andrea, F.

2017-12-01

We here investigate and test using large-eddy simulations the hypothesis that evaporative cooling might not be crucial for shallow clouds. Results from various Shallow convection and stratocumulus LES experiments show that the influence of evaporative cooling is secondary compared to turbulent mixing, which dominates the buoyancy reversal. In shallow cumulus subising shells are not due to evaporative cooling but rather reflect a vortical structure, with a postive buoyancy anomaly in the core due to condensation. Disabling evaporative cooling has negligible impact on this vortical structure and on buoyancy reversal. Similarly in non-precipitating stratocumuli evaporative cooling is negeligible copmared to other factors, especially turbulent mixing and pressure effects. These results emphasize that it may not be critical to icnlude evaporative cooling in parameterizations of shallow clouds and that it does not alter entrainment.

Physics experiment on the Dragon reactor

Energy Technology Data Exchange (ETDEWEB)

Hunt, C.

1974-10-15

The paper describes a set of DRAGON experiments planned to measure burn-up effects in DRAGON irradiated fuel. Irradiated fuel elements from DRAGON are to be subjected to reactivity measurements in the HECTOR experimental reactor to infer the residual U235 content followed by isotopic analyses at CEA laboratories in 1975. Fast neutron damage to DRAGON graphite is compared to fast neutron dose measurements using Ni58 (n,p) Co58 activation wires in both DRAGON and the DIDO MTR. Gamma scanning of irradiated fuel elements are used to compare axial power profiles to those derived from two-dimensional and three-dimensional calculations of the DRAGON reactor.

Sodium evaporation into a forced argon flow, (1)

International Nuclear Information System (INIS)

Kumada, Toshiaki; Kasahara, Fumio; Ishiguro, Ryoji

1976-01-01

Measurements were made on the rate of evaporation from a rectangular-shaped free surface of liquid sodium into argon flow. Tests were carried out at various levels of sodium temperature, of argon velocity and of argon temperature, under conditions where fog formation could be expected. To gain information on the enhancement of evaporation occasioned by fog formation, a supplementary experiment was performed on convection heat transfer into flowing air from a heated plate of the same geometry as the free surface of the sodium in the preceding measurements. The values obtained for the rate of evaporation and Sherwood number were compared with those predicted by the heat transfer experiment and by the theory by Hill and Szekely. The overall results revealed that the rate of sodium evaporation can amount to as much as three times that predicted by the heat transfer experiment, and that it varies roughly linearly with the heat transfer rate and with the sodium vapor pressure prevailing at the free surface. (auth.)

Decomposition of thermally unstable substances in film evaporators

Energy Technology Data Exchange (ETDEWEB)

Matz, G

1982-10-01

It is widely known that film evaporators are considered to permit really gentle evaporation of heat-sensitive substances. Nevertheless, decomposition of such substance still occurs to an extent depending upon the design and operation of the evaporator. In the following a distinction is made between evaporators with films not generated mechanically, namely the long tube evaporator (lTE) or climbing film evaporator, the falling film evaporator (FFE) and the multiple phase helical tube (MPT) or helical coil evaporators (TFE). Figs 1 and 2 illustrate the mode of operation. A theory of the decomposition of thermally unstable substances in these evaporators is briefly outlined and compared with measurements. Such a theory cannot be developed without any experimental checks; on the other hand, meausrements urgently need a theoretical basis if only to establish what actually has to be measured. All experiments are made with a system of readily adjustable decomposability, namely with aqueous solutions of saccharose; the thermal inversion of this compound can be controlled by addition of various amounts or concentrations of hydrochloric acid. In the absence of any catalysis by hydrochloric acid, the decomposition rates within in the temperature interval studied (60-130/sup 0/C) are so low that the experiments would take much too long and determination of the concentration differences (generally by polarimetric methods) would be very complicated. Such slight effects would also be very unfavourable for comparison with theory. (orig.)

Field evaporation test of uranium tailings solution

International Nuclear Information System (INIS)

Chandler, B.L.; Shepard, T.A.; Stewart, T.A.

1985-01-01

A field experiment was performed to observe the effect on evaporation rate of a uranium tailings impoundment pond water as salt concentration of the water increased. The duration of the experiment was long enough to cause maximum salt concentration of the water to be attained. The solution used in the experiment was tailings pond water from an inactive uranium tailings disposal site in the initial stages of reclamation. The solution was not neutralized. The initial pH was about 1.0 decreasing to a salt gel at the end of the test. The results of the field experiment show a gradual and slight decrease in evaporation efficiency. This resulted as salt concentrations increased and verified the practical effectiveness of evaporation as a water removal method. In addition, the physical and chemical nature of the residual salts suggest that no long-term stability problem would likely result due to their presence in the impoundment during or after reclamation

Dry cooling tower operating experience in the LOFT reactor

International Nuclear Information System (INIS)

Hunter, J.A.

1980-01-01

A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

Lessons from feedback of safety operating experience for reactor physics

International Nuclear Information System (INIS)

Suchomel, J.; Rapavy, S.

1999-01-01

Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

Experience from and research activities at the Otaniemi TRIGA reactor

International Nuclear Information System (INIS)

Bars, Bruno

1976-01-01

Experience from the Finnish TRIGA Reactor is reported, small changes and improvements in the control console of the Fir-1 reactor have been made. A minicomputer based data collecting system is planned and installed. It will be used for collecting data from operation and radiation monitors including the new isotope laboratory, and also simultaneously smaller experiments such as control rod calibration. A minicomputer is used for on-line reactor noise studies. The automatic uranium analyzer has a maximum sensitivity of 0.03 μg U 235 and 1.2 Th 232 . The system is now used at a sampling rate of about one sample per minute. (author)

Approximate computation of hydrothermal conditions of nuclear reactor spray ponds

International Nuclear Information System (INIS)

Yarkho, A.A.; Borshchev, V.A.

1990-01-01

An algorithm is presented for determining the evaporation numbers of nuclear reactor spray ponds which provide necessary reactor cooling during its normal operation under given meteorological conditions with account of restrictions on the cooled water temperature at the reactor entrance

Power cycling experiments in INR-TRIGA-SSR Reactor

International Nuclear Information System (INIS)

Dumitru, M.

2008-01-01

The in-reactor experimental program started this summer with some power cycling experiments to provide date on fuel behaviour under abnormal reactor operating conditions. The paper describes the irradiation device, its operational features and an original 'under-flux' movement system. Also, there are presented main data of irradiation device (pressure, flow, temperature, construction), in-pile section, location, sample, instrumentation, experimental sequences and operating data of Interest for the experimenters. (author)

Evaporating firewalls

Science.gov (United States)

Van Raamsdonk, Mark

2014-11-01

In this note, we begin by presenting an argument suggesting that large AdS black holes dual to typical high-energy pure states of a single holographic CFT must have some structure at the horizon, i.e. a fuzzball/firewall, unless the procedure to probe physics behind the horizon is state-dependent. By weakly coupling the CFT to an auxiliary system, such a black hole can be made to evaporate. In a case where the auxiliary system is a second identical CFT, it is possible (for specific initial states) that the system evolves to precisely the thermofield double state as the original black hole evaporates. In this case, the dual geometry should include the "late-time" part of the eternal AdS black hole spacetime which includes smooth spacetime behind the horizon of the original black hole. Thus, if a firewall is present initially, it evaporates. This provides a specific realization of the recent ideas of Maldacena and Susskind that the existence of smooth spacetime behind the horizon of an evaporating black hole can be enabled by maximal entanglement with a Hawking radiation system (in our case the second CFT) rather than prevented by it. For initial states which are not finely-tuned to produce the thermofield double state, the question of whether a late-time infalling observer experiences a firewall translates to a question about the gravity dual of a typical high-energy state of a two-CFT system.

Isotope effects accompanying evaporation of water from leaky containers.

Science.gov (United States)

Rozanski, Kazimierz; Chmura, Lukasz

2008-03-01

Laboratory experiments aimed at quantifying isotope effects associated with partial evaporation of water from leaky containers have been performed under three different settings: (i) evaporation into dry atmosphere, performed in a dynamic mode, (ii) evaporation into dry atmosphere, performed in a static mode, and (iii) evaporation into free laboratory atmosphere. The results demonstrate that evaporative enrichment of water stored in leaky containers can be properly described in the framework of the Craig-Gordon evaporation model. The key parameter controlling the degree of isotope enrichment is the remaining fraction of water in the leaking containers. Other factors such as temperature, relative humidity, or extent of kinetic fractionation play only minor roles. Satisfactory agreement between observed and predicted isotope enrichments for both (18)O and (2)H in experiments for the case of evaporation into dry atmosphere could be obtained only when molecular diffusivity ratios of isotope water molecules as suggested recently by Cappa et al. [J. Geophys. Res., 108, 4525-4535, (2003).] were adopted. However, the observed and modelled isotope enrichments for (2)H and (18)O could be reconciled also for the ratios of molecular diffusivities obtained by Merlivat [J. Chem. Phys., 69, 2864-2871 (1978).], if non-negligible transport resistance in the viscous liquid sub-layer adjacent to the evaporating surface is considered. The evaporation experiments revealed that the loss of mass of water stored in leaky containers in the order of 1%, will lead to an increase of the heavy isotope content in this water by ca. 0.35 and 1.1 per thousand, for delta (18)O and delta (2)H, respectively.

Visualization of steam bubbles with evaporation in molten alloy

International Nuclear Information System (INIS)

Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi; Takenaka, Nobuyuki; Matsubayashi, Masahito

1997-01-01

An innovative Steam Generator concept of Fast Breeder Reactors by using liquid-liquid direct contact heat transfer has been developed. In this concept, the SG shell is filled with a molten alloy heated by primary sodium. Water is fed into the high temperature molten alloy, and evaporates by direct contact heating. In order to obtain the fundamental information to discuss the heat transfer mechanisms of the direct contact between the water and the molten alloy, this phenomenon was visualized by neutron radiography. JRR-3M radiography in Japan Atomic Energy Research Institute was used. Followings are main results. (1) The bubbles with evaporation are risen with vigorous form changing, coalescence and break-up. Because of these vigorous evaporation, this system have the high heat transfer performance. (2) The rising velocities and volumes of bubbles are calculated from pixcel values of images. The velocities of the bubbles with evaporation are about 60 cm/s, which is larger than that of inert gas bubbles in molten alloy (20-40 cm/s). (3) The required heat transfer length of evaporation is calculated from pixcel values of images. The relation between heat transfer length and superheat temperature, obtained through the heat transfer test, is conformed by this calculation. (author)

Experiments with the SUR 100 training reactor

International Nuclear Information System (INIS)

Milicic, B.

1984-06-01

This paper contains a compilation of various experiments using the SUR - 100 reactor for training purposes, which have been widly proved in practical work at the School for Nuclear Technology of the Karlsruhe Research Center. (orig.) [de

Reactor physics experiment plan using TCA

Energy Technology Data Exchange (ETDEWEB)

Shimada, Shoichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-06-01

The Reduced-Moderation Water Reactor (RMWR) is one of the next generation water-cooled reactors, which aims at effective utilization of uranium resource, high burn-up, long operation cycle, and plutonium multi-recycle. For verification of the feasibility, negative void reactivity coefficient and conversion ratio more than 1.0 must be confirmed. This report is to plan critical experiments using TCA in JAERI. Critical Experiments performed so far in Europe and Japan are reviewed, and no useful data are available for RMWR development. Critical experiments using TCA (Tank Type Critical Assembly) in JAERI are planned. MOX fuel rods should be prepared for the experiments and some modifications of equipment are needed for use of MOX fuel rods. This report describes the preliminary plan of physics experiments. The number of MOX-fuel rods used in the experiments is obtained by calculations and modification of the equipment for the experiments are shown. New MOX fuel and UO{sub 2} fuel rods are necessary for the RMWR critical experiments. Number of MOX fuel rods is 1000 for Plutonium fissile enrichment of 5 wt%, 1000 for 10 wt%, 1500 for 15 wt% and 500 for 20 wt%, respectively. Depleted UO{sub 2} fuel rods for blanket/buffer region are 4000. Driver fuel rods of 4.9 wt% UO{sub 2} are 3000. Modification of TCA facility is requested to treat the large amount of MOX fuels from safety point of view. Additional shielding device at the top of the tank for loading the MOX fuels and additional safety plates to ensure safety are requested. The core is divided into two regions by inserting an inner tank to avoid criticality in MOX region only. The test region is composed by MOX fuel rods in the inner tank. Criticality is established by UO{sub 2} driver fuel rods outside of the inner tank. (Tsuchihashi, K.)

Designing a new highly active liquid evaporator - 16075

International Nuclear Information System (INIS)

Robson, Paul; Candy, Emma

2009-01-01

The Highly Active Liquid Effluent Storage (HALES) plant stores, concentrates and conditions Highly Active Liquor (HAL) in evaporators for buffer storage in Highly Active Storage Tanks (HAST). Highly Active (HA) evaporators play a pivotal role in the delivery of reprocessing, historic clean up and hazard reduction missions across the Sellafield site. In addition to the engineering projects implemented to extend the life expectation of the current evaporator fleet, the UK Nuclear Decommissioning Agency (NDA) is sponsoring the construction of a new HA evaporator (Evaporator D) on the Sellafield site. The design and operation of the new HA evaporator is based on existing/recent HA evaporator technology but learning from past operational experience. Operational experience has been a key area where the existing plant operators have influenced both the new design itself and the requirements for commissioning and training. Many of the learning experiences require relatively simple engineering design modifications such as a new internal washing provision and transfer line blockage recovery systems, they are never-the-less expected to significantly improve the flexibility and operational capability of the new evaporator. Issues that the project delivery team has addressed as part of the development of the design and construction have included: - Minimising interruptions and/or changes to the normal operations of interfacing plants during construction, commissioning and operation of the new facility. - Modularization of the plant, enabling fabrication of the majority of the plant equipment off-site within a workshop (as opposed to on-site) environment improving Quality Assurance and reducing on-Site testing needs. - Drawing out the balance between operational and corrosion resistance improvements with actual design and delivery needs. - Provision of a new facility reliant on the infrastructure of an existing and ageing facility and the competing demands of the related safety

Predictions on an HTR coolant composition after operational experience with experimental reactors

International Nuclear Information System (INIS)

Nieder, R.

1981-01-01

Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

Influence of Evaporation on Soap Film Rupture.

Science.gov (United States)

Champougny, Lorène; Miguet, Jonas; Henaff, Robin; Restagno, Frédéric; Boulogne, François; Rio, Emmanuelle

2018-03-13

Although soap films are prone to evaporate due to their large surface to volume ratio, the effect of evaporation on macroscopic film features has often been disregarded in the literature. In this work, we experimentally investigate the influence of environmental humidity on soap film stability. An original experiment allows to measure both the maximum length of a film pulled at constant velocity and its thinning dynamics in a controlled atmosphere for various values of the relative humidity [Formula: see text]. At first order, the environmental humidity seems to have almost no impact on most of the film thinning dynamics. However, we find that the film length at rupture increases continuously with [Formula: see text]. To rationalize our observations, we propose that film bursting occurs when the thinning due to evaporation becomes comparable to the thinning due to liquid drainage. This rupture criterion turns out to be in reasonable agreement with an estimation of the evaporation rate in our experiment.

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

International Nuclear Information System (INIS)

Moiseyev, A.V.

2008-01-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

2008-07-01

There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

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

International Nuclear Information System (INIS)

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

2000-06-01

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

Physics experiments with the operating reactor

Energy Technology Data Exchange (ETDEWEB)

Cullington, G R; King, D C

1973-09-27

Experimental techniques have been developed and used on Dragon to give consistent information on excess reactivity and shut down margin. The reactivity measurements have been correlated with the theoretical calculations and have led to improvements in the calculations. The methods used and the results obtained are accepted by the Safety Committee as sufficient evidence for compliance with the fuel loading safety rules. Although the reactor was not designed as an experimental facility, flux and dose measurements experiments have been successfully carried out. Mass flow and negative reactivity transient measurements have been carried out. These are valuable for demonstration of the flexibility of the reactor system and for giving confidence in theoretical calculations.

Propagation calculation for reactor cases

Energy Technology Data Exchange (ETDEWEB)

Yang Yanhua [School of Power and Energy Engineering, Shanghai Jiao Tong Univ., Shanghai (China); Moriyama, K.; Maruyama, Y.; Nakamura, H.; Hashimoto, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2000-11-01

The propagation of steam explosion for real reactor geometry and conditions are investigated by using the computer code JASMINE-pro. The ex-vessel steam explosion is considered, which is described as follow: during the accident of reactor core meltdown, the molten core melts a hole at the bottom of reactor vessel and causes the higher temperature core fuel being leaked into the water pool below reactor vessel. During the melt-water mixing interaction process, the high temperature melt evaporates the cool water at an extreme high rate and might induce a steam explosion. A steam explosion could experience first the premixing phase and then the propagation explosion phase. For a propagation calculation, we should know the information about the initial fragmentation time, the total melt mass, premixing region size, initial void fraction and distribution of the melt volume fraction, and so on. All the initial conditions used in this calculation are based on analyses from some simple assumptions and the observation from the experiments. The results show that the most important parameter for the initial condition of this phase is the total mass and its initial distribution. This gives the requirement for a premixing calculation. On the other hand, for higher melt volume fraction case, the fragmentation is strong so that the local pressure can exceed over the EOS maximum pressure of the code, which lead to the incorrect calculation or divergence of the calculation. (Suetake, M.)

Tritium concentration reducing method in atmosphere in nuclear reactor containment facility

International Nuclear Information System (INIS)

Hirasawa, Yoshiya; Kigoshi, Yasutane; Yonenaga, Haruo.

1992-01-01

A portion of water content in an atmosphere is condensed by a condensation/evaporation device disposed in a nuclear reactor containment building and then a portion of the condensed water is evaporated in the atmosphere. A portion of hydrogen nuclides constituting the evaporated water content is subjected to isotopic exchange with tritium nuclides in the atmosphere. A portion of water content in the atmosphere applied with the isotopic exchange is condensed in the condensation/evaporation device. That is, the hydrogen nuclides in steams are applied with isotopic exchange with tritium nuclides, and steams incorporating tritium nuclides are condensed again in the condensation/evaporation device, to transfer the tritium nuclides in the atmosphere to condensed water. The condensed water is recovered without releasing the tritium nuclides to the outside of the reactor containment building, thereby enabling to reduce the tritium concentration in the atmosphere. (N.H.)

Behavior of pressure rise and condensation caused by water evaporation under vacuum at high temperature

International Nuclear Information System (INIS)

Takase, Kazuyuki; Kunugi, Tomoaki; Yamazaki, Seiichiro; Fujii, Sadao

1998-01-01

Pressure rise and condensation characteristics during the ingress-of-coolant event (ICE) in fusion reactors were investigated using the preliminary ICE apparatus with a vacuum vessel (VV), an additional tank (AT) and an isolation valve (IV). A surface of the AT was cooled by water at RT. The high temperature and pressure water was injected into the VV which was heated up to 250degC and pressure and temperature transients in the VV were measured. The pressure increased rapidly with an injection time of the water because of the water evaporation. After the IV was opened and the VV was connected with the AT, the pressure in the VV decreased suddenly. From a series of the experiments, it was confirmed that control factors on the pressure rise were the flushing evaporation and boiling heat transfer in the VV, and then, condensation of the vapor after was effective to the depressurization in the VV. (author)

The double chooz reactor neutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Botella, I Gil [CIEMAT, Basic Research Department, Avenida Complutense, 22, 28040 Madrid (Spain)], E-mail: ines.gil@ciemat.es

2008-05-15

The Double Chooz reactor neutrino experiment will be the next detector to search for a non vanishing {theta}{sub 13} mixing angle with unprecedented sensitivity, which might open the way to unveiling CP violation in the leptonic sector. The measurement of this angle will be based in a precise comparison of the antineutrino spectrum at two identical detectors located at different distances from the Chooz nuclear reactor cores in France. Double Chooz is particularly attractive because of its capability to measure sin{sup 2} (2{theta}{sub 13}) to 3{sigma} if sin{sup 2}(2{theta}{sub 13}) > 0.05 or to exclude sin{sup 2}(2{theta}{sub 13}) down to 0.03 at 90% C.L. for {delta}m{sup 2} = 2.5 x 10{sup -3} eV{sup 2} in three years of data taking with both detectors. The construction of the far detector starts in 2008 and the first neutrino results are expected in 2009. The current status of the experiment, its physics potential and design and expected performance of the detector are reviewed.

Substantiation of physical concepts of fast reactors in Russia: experience and prospects

Energy Technology Data Exchange (ETDEWEB)

Alekseev, P.N. [Russian Research Center ' Kurchatov Institute' (RRC KI), 1, Kurchatov Sq., Moscow, 123182 (Russian Federation); Vasiliev, B.A. [Experimental Design Bureau of Machine Building (OKBM) 15, Burnakovskiy Pr., N. Novgorod, 603074 (Russian Federation); Kormilitsyn, M.V. [State Scientific Center of Russian Federation - Research Institute of Atomic Reactors (NIIAR) Dimitrovgrad-10, Ulianovsk Reg., 433510 (Russian Federation); Lopatkin, A.V. [N.A. Dollezhal Research and Development Institute of Power Engineering (NIKIET) 2/8, M. Krasnoselskaya Str., Moscow, 107140 (Russian Federation); Seleznev, E.F. [All-Russian Research Institute for Nuclear Power Plant Operation (VNIIAES) 25, Ferganskaya, Moscow, 109507 (Russian Federation); Khomyakov, Yu.S.; Tsybulia, A.M. [State Scientific Center of the Russian Federation - A. I. Leypunsky Institute for Physics and Power Engineering (SSC RF- IPPE) 1, Bondarenko Sq., Obninsk, Kaluga Reg., 249033 (Russian Federation); Tocheny, L.V. [International Science and Technology Center (ISTC) 32-34 Krasnoproletarskaya Ulitsa, Moscow, 127473 (Russian Federation)

2008-07-01

The fast reactor concept in Russia has accumulated unique experience, since its advent in the 1950's and up to the present, from the creation of the first experimental installation BR-1, experimental reactors BR-5 and BOR-60, the pilot industrial reactors BN-350 in Kazakhstan and up to the BN-600 at Beloyarsk Atomic Power Station. Investigations on the first experimental installations BR-1 and BR-5/-10 proved the propriety of the idea that it is possible to create nuclear reactors that can produce more nuclear fuel than they consume, i.e. the idea of breeding. The architecture of such reactors was also designed, producing a current leader among fast reactors with sodium coolant and oxide uranium-plutonium fuel. Operational experience of BOR-60, BN-350 and, particularly, BN-600 confirmed the engineering and technical feasibility of the concept of fast reactors, the possibility for its realization both for power production and for certain other purposes as well, such as desalinisation of sea water (BN-350) and for radionuclide production (BN-350, BN-600), and it enabled the development and verification of different models, computer methods and codes. The paper presents a review of experience in the creation of plants with fast reactors, scientific research on these installations, principal results, the current status of experimental data analysis, and prospective directions in the development of fast reactors and the corresponding experimental basis in Russia. (authors)

U-233 fuelled low critical mass solution reactor experiment PURNIMA II

International Nuclear Information System (INIS)

Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

1987-01-01

A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

Water evaporation in silica colloidal deposits.

Science.gov (United States)

Peixinho, Jorge; Lefèvre, Grégory; Coudert, François-Xavier; Hurisse, Olivier

2013-10-15

The results of an experimental study on the evaporation and boiling of water confined in the pores of deposits made of mono-dispersed silica colloidal micro-spheres are reported. The deposits are studied using scanning electron microscopy, adsorption of nitrogen, and adsorption of water through attenuated total reflection-infrared spectroscopy. The evaporation is characterized using differential scanning calorimetry and thermal gravimetric analysis. Optical microscopy is used to observe the patterns on the deposits after evaporation. When heating at a constant rate and above boiling temperature, the release of water out of the deposits is a two step process. The first step is due to the evaporation and boiling of the surrounding and bulk water and the second is due to the desorption of water from the pores. Additional experiments on the evaporation of water from membranes having cylindrical pores and of heptane from silica deposits suggest that the second step is due to the morphology of the deposits. Copyright © 2013 Elsevier Inc. All rights reserved.

Water evaporation: a transition path sampling study.

Science.gov (United States)

Varilly, Patrick; Chandler, David

2013-02-07

We use transition path sampling to study evaporation in the SPC/E model of liquid water. On the basis of thousands of evaporation trajectories, we characterize the members of the transition state ensemble (TSE), which exhibit a liquid-vapor interface with predominantly negative mean curvature at the site of evaporation. We also find that after evaporation is complete, the distributions of translational and angular momenta of the evaporated water are Maxwellian with a temperature equal to that of the liquid. To characterize the evaporation trajectories in their entirety, we find that it suffices to project them onto just two coordinates: the distance of the evaporating molecule to the instantaneous liquid-vapor interface and the velocity of the water along the average interface normal. In this projected space, we find that the TSE is well-captured by a simple model of ballistic escape from a deep potential well, with no additional barrier to evaporation beyond the cohesive strength of the liquid. Equivalently, they are consistent with a near-unity probability for a water molecule impinging upon a liquid droplet to condense. These results agree with previous simulations and with some, but not all, recent experiments.

Study on the Neutrino Oscillation with a Next Generation Medium-Baseline Reactor Experiment

International Nuclear Information System (INIS)

Joo, Kyung Kwang; Shin, Chang Dong

2014-01-01

For over fifty years, reactor experiments have played an important role in neutrino physics, in both discoveries and precision measurements. One of the methods to verify the existence of neutrino is the observation of neutrino oscillation phenomena. Electron antineutrinos emitted from a reactor provide the measurement of the small mixing angle θ 13 , providing rich programs of neutrino properties, detector development, nuclear monitoring, and application. Using reactor neutrinos, future reactor neutrino experiments, more precise measurements of θ 12 ,Δm 12 2 , and mass hierarchy will be explored. The precise measurement of θ 13 would be crucial for measuring the CP violation parameters at accelerators. Therefore, reactor neutrino physics will assist in the complete understanding of the fundamental nature and implications of neutrino masses and mixing. In this paper, we investigated several characteristics of RENO-50, which is a future medium-baseline reactor neutrino oscillation experiment, by using the GloBES simulation package

Operating experience feedback from safety significant events at research reactors

Energy Technology Data Exchange (ETDEWEB)

Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

2015-05-15

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Operating experience feedback from safety significant events at research reactors

International Nuclear Information System (INIS)

Shokr, A.M.

2015-01-01

Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

Evaporative Cooling of Antiprotons to Cryogenic Temperatures

CERN Document Server

Andresen, G B; Baquero-Ruiz, M; Bertsche, W; Bowe, P D; Butler, E; Cesar, C L; Chapman, S; Charlton, M; Fajans, J; Friesen, T; Fujiwara, M C; Gill, D R; Hangst, J S; Hardy, W N; Hayano, R S; Hayden, M E; Humphries, A; Hydomako, R; Jonsell, S; Kurchaninov, L; Lambo, R; Madsen, N; Menary, S; Nolan, P; Olchanski, K; Olin, A; Povilus, A; Pusa, P; Robicheaux, F; Sarid, E; Silveira, D M; So, C; Storey, J W; Thompson, R I; van der Werf, D P; Wilding, D; Wurtele, J S; Yamazaki, Y

2010-01-01

We report the application of evaporative cooling to clouds of trapped antiprotons, resulting in plasmas with measured temperature as low as 9~K. We have modeled the evaporation process for charged particles using appropriate rate equations. Good agreement between experiment and theory is observed, permitting prediction of cooling efficiency in future experiments. The technique opens up new possibilities for cooling of trapped ions and is of particular interest in antiproton physics, where a precise CPT test on trapped antihydrogen is a long-standing goal.

Numerical investigation of liquid methanol evaporation and oxy-combustion inside a button-cell ITM reactor

International Nuclear Information System (INIS)

Nemitallah, Medhat A.; Habib, Mohamed A.

2017-01-01

Highlights: • Analysis of liquid methanol evaporation and oxy-combustion in an ITM reactor. • A semi-empirical model is applied after fitting with the available LNO membrane data. • Influences of inlet fuel fraction, inlet gas temperature and inlet sweep flux are studied. • High combustion efficiency is encountered at moderate inlet gas temperatures. • High fuel concentration at low inlet sweep flow resulted in high oxygen flux. - Abstract: A numerical study is conducted to investigate the performance of a button-cell LNO-ITM reactor utilizing the soot-free oxygenated liquid methanol under oxy-combustion condition. The Euler-Lagrange approach is utilized to solve discrete phase model. Taylor analogy breakup (TAB) model is used due to its convenience with the cases of low injection speed. A plain orifice atomizer is used for fuel atomization and CO_2 is used as sweep gas. A semi-empirical oxygen permeation model (ABn model) is validated with the available experimental data and is, then, applied in the present model. Over a wide range of inlet fuel concentrations, the results showed increase in oxygen permeation flux of about five times in cases of reacting conditions as compared to the cases of non-reacting cases. The results showed high oxygen permeation flux at low inlet fuel concentrations due to the improvement in the oxygen to fuel ratio toward the stoichiometric conditions. At inlet gas temperatures of 1223 K, 1123 K, 1023 K and 923 K, the combustion temperature approached 1423 K, 1347 K, 1284 K and 1231 K, respectively, indicating an average combustion efficiency of 43% at moderate inlet gas temperatures. High fuel concentration at low inlet sweep flow resulted in high oxygen flux and high combustion temperature.

Liquid metal cooled reactors: Experience in design and operation

International Nuclear Information System (INIS)

2007-12-01

on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

International Nuclear Information System (INIS)

Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

2015-01-01

Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

Calculation of vapor pressures of oxide fuels up to 5,000 K for equilibrium and nonequilibrium evaporation

International Nuclear Information System (INIS)

Breitung, W.

1975-06-01

In the first part of this work the evaporation kinetics of multicomponent systems is studied with UO 2 as the example. The evaporation, which is generally incongruent, implies that two opposing types of steady-state evaporation must be distinguished: equilibrium evaporation and 'forced congruent' evaporation. The two types of evaporation indicated entail different vapor pressures. In some prompt critical reactor incidents forced congruent evaporation must be anticipated. The second part of this work contains the calculation of the vapor pressures of UOsub(2+-x) and (U,Pu)Osub(2+-x) for both types of evaporation up to temperature of 5,000 K. The calculating procedures are based on the method of Rand and Markin (1967) incorporating the recent thermodynamic data. The agreement between the measured and calculated total pressures is good for the ranges of temperature and stoichiometry for which experimental results are available. This supports the results calculated for higher temperature ranges. (orig./UA) [de

The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

International Nuclear Information System (INIS)

Junkrans, S.; Vesterlund, G.; Vaernild, O.

1980-01-01

The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

International Nuclear Information System (INIS)

Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

2011-01-01

The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

A modified surface-resistance approach for representing bare-soil evaporation: wind tunnel experiments under various atmospheric conditions

International Nuclear Information System (INIS)

Yamanaka, T.; Takeda, A.; Sugita, F.

1997-01-01

A physically based (i.e., nonempirical) representation of surface-moisture availability is proposed, and its applicability is investigated. This method is based on the surface-resistance approaches, and it uses the depth of evaporating surface rather than the water content of the surface soil as the determining factor of surface-moisture availability. A simple energy-balance model including this representation is developed and tested against wind tunnel experiments under various atmospheric conditions. This model can estimate not only the latent heat flux but also the depth of the evaporating surface simultaneously by solving the inverse problem of energy balance at both the soil surface and the evaporating surface. It was found that the depth of the evaporating surface and the latent heat flux estimated by the model agreed well with those observed. The agreements were commonly found out under different atmospheric conditions. The only limitation of this representation is that it is not valid under conditions of drastic change in the radiation input, owing to the influence of transient phase transition of water in the dry surface layer. The main advantage of the approach proposed is that it can determine the surface moisture availability on the basis of the basic properties of soils instead of empirical fitting, although further investigations on its practical use are needed

Water Evaporation in Swimming Baths

DEFF Research Database (Denmark)

Hyldgård, Carl-Erik

This paper is publishing measuring results from models and full-scale baths of the evaporation in swimming baths, both public baths and retraining baths. Moreover, the heat balance of the basin water is measured. In addition the full-scale measurements have given many experiences which are repres......This paper is publishing measuring results from models and full-scale baths of the evaporation in swimming baths, both public baths and retraining baths. Moreover, the heat balance of the basin water is measured. In addition the full-scale measurements have given many experiences which...... are represented in instructions for carrying out and running swimming baths. If you follow the instructions you can achieve less investments, less heat consumption and a better comfort to the bathers....

Human Factors Engineering (HFE) insights for advanced reactors based upon operating experience

International Nuclear Information System (INIS)

Higgins, J.; Nasta, K.

1997-01-01

The NRC Human Factors Engineering Program Review Model (HFE PRM, NUREG-0711) was developed to support a design process review for advanced reactor design certification under 10CFR52. The HFE PRM defines ten fundamental elements of a human factors engineering program. An Operating Experience Review (OER) is one of these elements. The main purpose of an OER is to identify potential safety issues from operating plant experience and ensure that they are addressed in a new design. Broad-based experience reviews have typically been performed in the past by reactor designers. For the HFE PRM the intent is to have a more focussed OER that concentrates on HFE issues or experience that would be relevant to the human-system interface (HSI) design process for new advanced reactors. This document provides a detailed list of HFE-relevant operating experience pertinent to the HSI design process for advanced nuclear power plants. This document is intended to be used by NRC reviewers as part of the HFE PRM review process in determining the completeness of an OER performed by an applicant for advanced reactor design certification. 49 refs

Experience with valves for PHWR reactors

International Nuclear Information System (INIS)

Narayan, K.; Mhetre, S.G.

1977-01-01

Material specifications and inspection and testing requirements of the valves meant for use in nuclear reactors are mentioned. In the heavy water systems (both primary and moderator) of a PHWR type reactor, the valves used are gate valves, globe valves, diaphragm valves, butterfly valves, check valves and relief valves. Their locations and functions they perform in the Rajasthan Atomic Power Station Unit-1 are described. Experience with them is given. The major problems encountered with them have been : (1) leakage from the stem seals and body bonnet joint, (2) leakage due to failure of diaphragm and/or washout of the packing and (3) malfunctioning. Measures taken to solve these are discussed. Finally a mention has been made of improved versions of valves, namely, metal diaphragm valve and inverted relief valve. (M.G.B.)

US graphite reactor D ampersand D experience

International Nuclear Information System (INIS)

Garrett, S.M.K.; Williams, N.C.

1997-02-01

This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE)

Thermal-Hydraulic Experiments and Modelling for Advanced Nuclear Reactor Systems

International Nuclear Information System (INIS)

Song, C. H.; Baek, W. P.; Chung, M. K.

2007-06-01

The objectives of the project are to study thermal hydraulic characteristics of advanced nuclear reactor system for evaluating key thermal-hydraulic phenomena relevant to new safety concepts. To meet the research goal, several thermal hydraulic experiments were performed and related thermal hydraulic models were developed with the experimental data which were produced through the thermal hydraulic experiments. The Followings are main research topics: - Multi-dimensional Phenomena in a Reactor Vessel Downcomer - Condensation-induced Thermal Mixing in a Pool - Development of Thermal-Hydraulic Models for Two-Phase Flow - Construction of T-H Data Base

Nuclear data and integral experiments in reactor physics

International Nuclear Information System (INIS)

Farinelli, U.

1980-01-01

The material given here broadly covers the content of the 10 lectures delivered at the Winter Course on Reactor Theory and Power Reactors, ICTP, Trieste (13 February - 10 March 1978). However, the parts that could easily be found in the current literature have been omitted and replaced with the appropriate references. The needs for reactor physics calculations, particularly as applicable to commercial reactors, are reviewed in the introduction. The relative merits and shortcomings of fundamental and semi-empirical methods are discussed. The relative importance of different nuclear data, the ways in which they can be measured or calculated, and the sources of information on measured and evaluated data are briefly reviewed. The various approaches to the condensation of nuclear data to multigroup cross sections are described. After some consideration to the sensitivity calculations and the evaluation of errors, some of the most important type of integral experiments in reactor physics are introduced, with a view to showing the main difficulties in the interpretation and utilization of their results and the most recent trends in experimentation. The conclusions try to assign some priorities in the implementation of experimental and calculational capabilities, especially for a developing country. (author)

How internal drainage affects evaporation dynamics from soil surfaces ?

Science.gov (United States)

Or, D.; Lehmann, P.; Sommer, M.

2017-12-01

Following rainfall, infiltrated water may be redistributed internally to larger depths or lost to the atmosphere by evaporation (and by plant uptake from depths at longer time scales). A large fraction of evaporative losses from terrestrial surfaces occurs during stage1 evaporation during which phase change occurs at the wet surface supplied by capillary flow from the soil. Recent studies have shown existence of a soil-dependent characteristic length below which capillary continuity is disrupted and a drastic shift to slower stage 2 evaporation ensues. Internal drainage hastens this transition and affect evaporative losses. To predict the transition to stage 2 and associated evaporative losses, we developed an analytical solution for evaporation dynamics with concurrent internal drainage. Expectedly, evaporative losses are suppressed when drainage is considered to different degrees depending on soil type and wetness. We observe that high initial water content supports rapid drainage and thus promotes the sheltering of soil water below the evaporation depth. The solution and laboratory experiments confirm nonlinear relationship between initial water content and total evaporative losses. The concept contributes to establishing bounds on regional surface evaporation considering rainfall characteristics and soil types.

Miniature electron bombardment evaporation source: evaporation rate measurement

International Nuclear Information System (INIS)

Nehasil, V.; Masek, K.; Matolin, V.; Moreau, O.

1997-01-01

Miniature electron beam evaporation sources which operate on the principle of vaporization of source material, in the form of a tip, by electron bombardment are produced by several companies specialized in UHV equipment. These sources are used primarily for materials that are normally difficult to deposit due to their high evaporation temperature. They are appropriate for special applications such as heteroepitaxial thin film growth requiring a very low and well controlled deposition rate. A simple and easily applicable method of evaporation rate control is proposed. The method is based on the measurement of ion current produced by electron bombardment of evaporated atoms. The absolute evaporation flux values were measured by means of the Bayard-Alpert ion gauge, which enabled the ion current vs evaporation flux calibration curves to be plotted. (author). 1 tab., 4 figs., 6 refs

Decommissioning experience of the Japan power demonstration reactor

International Nuclear Information System (INIS)

Hoshi, T.; Yanagihara, S.; Tachibana, M.; Momma, T.

1992-01-01

Actual dismantling of the Japan Power Demonstration Reactor (JPDR) has been progressing since 1986 aiming to make stage 3 condition as the final goal. Such highly activated components as the reactor pressure vessel (RPV) and the inner portion of biological shield concrete close to the RPV have removed using the remotely operated cutting machines. Useful data on the dismantling techniques and their safety as well as the manpower expenditure and radiation exposure of workers have been obtained. Experiences gained through the dismantling works are described in this paper. (author)

Theory and experiments on electrohydrodynamic enhancement of evaporation from water drops

International Nuclear Information System (INIS)

Barthakur, N.N.

1990-01-01

Space charge produce by a single corona electrode was used to enhance evaporation rates from sessile drops of water. The drying curve was traced and a drop lifetime determined by a beta ray gauge which provided both sensitivity and reproducibility to the measurements. Lifetime was reduced by a factor of 3.5 to 4.7 when subjected to fluxes of 3.02x10 12 positive charges cm -2 s -1 than those from freely evaporating drops in the laboratory. A theoretical model based on mass transfer coefficient was developed to predict the drop lifetime. Calculated lifetime of drops of volume 0.1 to 0.5 ml agreed within 12 percent of the experimental values. Electric wind caused by the ionic drag is proposed to be the principal driving force for the observed enhancement of evaporation from the drops. (author). 24 refs., 2 figs., 1 tab

Chasing {theta}{sub 13} with new reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Lasserre, Th. [DSM/DAPNIA/SPP, CEA/Saclay, 91191 Gif-sur-Yvette (France)

2005-12-15

It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the {theta}{sub 13} mixing angle, free from any parameter degeneracies and correlations induced by matter effect and the unknown leptonic Dirac CP phase. The current best constraint on the third mixing angle comes from the Chooz reactor neutrino experiment sin{sup 2}(2{theta}{sub 13})<0.2 (90 % C.L., {delta}m{sub atm}{sup 2}=2.010{sup -3} eV{sup 2}). Several projects of experiment, with different timescales, have been proposed over the last two years all around the world. Their sensitivities range from sin{sup 2}(2{theta}{sub 13})<0.01 to 0.03, having thus an excellent discovery potential of the {nu}{sub e} fraction of {nu}{sub 3}.

Daya bay reactor neutrino experiment

International Nuclear Information System (INIS)

Cao Jun

2010-01-01

Daya Bay Reactor Neutrino Experiment is a large international collaboration experiment under construction. The experiment aims to precisely determine the neutrino mixing angle θ 13 by detecting the neutrinos produced by the Daya Bay Nuclear Power Plant. θ 13 is one of two unknown fundamental parameters in neutrino mixing. Its magnitude is a roadmap of the future neutrino physics, and very likely related to the puzzle of missing antimatter in our universe. The precise measurement has very important physics significance. The detectors of Daya Bay is under construction now. The full operation is expected in 2011. Three years' data taking will reach the designed the precision, to determine sin 2 2θ 13 to better than 0.01. Daya Bay neutrino detector is an underground large nuclear detector of low background, low energy, and high precision. In this paper, the layout of the experiment, the design and fabrication progress of the detectors, and some highlighted nuclear detecting techniques developed in the detector R and D are introduced. (author)

300 area solvent evaporator interim status closure plan: Revision 2

International Nuclear Information System (INIS)

1989-02-01

This document describes activities for the closure of a hazardous waste tank treatment facility operated by the US Department of Energy-Richland Operations Office (DOE-RL) and co-operated by the Westinghouse Hanford Company (WHC). This treatment facility was a solvent evaporator located in the 300 Area of the Hanford Site, from 1975 to 1985 on behalf of DOE-RL. The 300 Area Solvent Evaporator (300 ASE) was a modified load lugger (dumpster) in which solvent wastes were evaporated. Some of the solvents were radioactively contaminated because they came from a degreaser which processed bare uranium metal billets from the N Reactor Fuel Manufacturing Facility. The waste was composed of perchloroethylene, trichloroethylene, 1,1,1-trichloroethane, ethyl acetate/bromine solution, paint shop solvents and possibly some used oil. Also, small amounts of uranium, copper, zirconium and possibly beryllium were present in the degreaser solvents as particulates. Radioactive and non-radioactive solvents were not segregated in the 300 ASE, and the entire mixture was regarded as mixed waste

Operating Experience from Events Reported to the IAEA Incident Reporting System for Research Reactors

International Nuclear Information System (INIS)

2015-03-01

Operating experience feedback is an effective mechanism in providing lessons learned from events and the associated corrective actions to prevent them, helping to improve safety at nuclear installations. The Incident Reporting System for Research Reactors (IRSRR), which is operated by the IAEA, is an important tool for international exchange of operating experience feedback for research reactors. The IRSRR reports contain information on events of safety significance with their root causes and lessons learned which help in reducing the occurrence of similar events at research reactors. To improve the effectiveness of the system, it is essential that national organizations demonstrate an appropriate interest for the timely reporting of events important to safety and share the information in the IRSRR database. At their biennial technical meetings, the IRSRR national coordinators recommended collecting the operating experience from the events reported to the IRSRR and disseminating it in an IAEA publication. This publication highlights the root causes, safety significance, lessons learned, corrective actions and the causal factors for the events reported to the IRSRR up to September 2014. The publication also contains relevant summary information on research reactor events from sources other than the IRSRR, operating experience feedback from the International Reporting System for Operating Experience considered relevant to research reactors, and a description of the elements of an operating experience programme as established by the IAEA safety standards. This publication will be of use to research reactor operating organizations, regulators and designers, and any other organizations or individuals involved in the safety of research reactors

Study on parameters of self-oscillations of the coolant flow rate in an evaporating channel of a boiling-type reactor

International Nuclear Information System (INIS)

Proshutinskij, A.P.; Lobachev, A.G.

1979-01-01

The experimental data on the oscillation frequencies and amplitudes of the coolant flow rate at the limit of the thermohydraulic stability of the boiling type reactor evaporating channel are presented. The experiments have been carried out on the channel simulators of three modifications -smooth-tube, with intensifiers of a transverse crimp type and of an inner spiral ribbing type. The range of the investigated regime parameters is as follows: the pressure - 2.5-14MPa; the heat flux density is 0.015-0.8MV/m 2 , mass velocity is 252-2520 kg/(m 2 xs), the temperature at the channel entrance is from 50 deg C up to (tsub(s) -5)deg C. The experimental data analysis is carried out on the assumption that the period of parameter oscillations in the steam generating channel equals the time of the coolant transfer through the channel. The formular is obtained which provides 25% accuracy of the oscillation frequency calculation in the range of underheating parameter variation B=0.5-3.0. As a result the following conclusions have been made: the oscillation frequency of the coolant flow rate is connected with the time of its transfer through the channel and does not practically depend on the type of the heat exchange intensifiers and the degree of the flux throttling at the channel entrance; the self-oscillation amplitude of the coolant flow rate depends on the regime and structural parameters as well

The review of the reactor physics experiments carried out on the LR-0 research reactor NRI Rez plc for reactors of the VVER type

International Nuclear Information System (INIS)

Hudec, Frantisek; Jansky, Bohumil; Juricek, Vlastimil; Mikus, Jan; Novak, Evzen; Osmera, Bohumil; Posta, Severin; Rypar, Vojtech; Svadlenkova, Marie

2010-01-01

LR-0 is an experimental zero power reactor mainly used for the determination of the neutron-physical characteristics of WWER and PWR type reactor lattices and shielding with UO2 or MOX fuel. Its major assets include capability to design and operate multizone cores, i.e. substituted cores, with an inner inserted part in hexagonal or square geometry (driven by LR-0 standard assemblies); Standard and special supporting plates for mock-up experiments; special supporting plates, which enables the triangular symmetrical assembly arrangement with an arbitrary pitch; Modeling neutron field parameters of power reactors; Wide range benchmarking possibilities, with high reproducibility of the benchmark design parameters; Wide range of measurement techniques including equipment and experienced personal; Flexible rearrangements of the core. The main experiments included: Pin wise flux distribution measurements; VVER-440 and VVER-1000 mock-ups; compact spent fuel storage; space kinetics experiment; core parameters experimental determination; experiment with new design fuel assembly; WWER-440 control assembly influence; and burnable absorber study. International research projects are also described. (P.A.)

An Overview of the International Reactor Physics Experiment Evaluation Project

International Nuclear Information System (INIS)

Briggs, J. Blair; Gulliford, Jim

2014-01-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

Fast breeder reactors: Experience and trends. V. 2

International Nuclear Information System (INIS)

1986-01-01

The IAEA Symposium on ''Fast Breeder Reactors: Experience and Future Trends'' was held, at the invitation of the Government of France, in Lyons, France, on 22-26 July 1985. It was hosted by the French Commissariat a l'energie atomique and Electricite de France. The purpose of the Symposium was to review the experience gained so far in the field of LMFBRs, taking into account the constructional, operational, technological, economic and fuel cycle aspects, and to consider the developmental trends as well as the international co-operation in fast breeder reactor design and utilization. The Symposium was attended by almost 400 participants (340 participants, 35 observers and 20 journalists) from 25 countries and five international organizations. More than 80 papers were presented and discussed during six regular sessions and four poster sessions. A separate abstract was prepared for each of these papers

Air conditioning device for reactor buildings

International Nuclear Information System (INIS)

Kikuchi, Shiro.

1982-01-01

Purpose: To decrease the opening areas of pipe lines for an air conditioning device at the portions passing through the shielding walls of a reactor building for a FBR type reactor, as well as reduce the size of the building. Constitution: Airs in the building for containing reactor are liquefied in an air liquefying mechanism. The liquefied airs are sent by way of pipe lines to each of evaporators, wherein each of the chambers are cooled because of latent heat of evaporation and evaporated airs are released to each of the chambers. The airs released to each of the chambers are collected into an exhaust chamber and sent by way of a duct to the air liquefying mechanism and liquefied again. Since the volume of the liquefied airs may be smaller than the amount conventionally required for usual cooled airs, the pipe lines passing through the shielding walls of the building can be of smaller diameter. This can decrease the opening areas of the pipe lines at the portions passing through the walls of the shieldings and, since the opening areas are smaller, the structure of the radiation shieldings can be simplified in these portions. Further, since the space of the pipe lines in the building is reduced extremely, the size of the building can be reduced. (Moriyama, K.)

Evaporation From Soil Containers With Irregular Shapes

Science.gov (United States)

Assouline, Shmuel; Narkis, Kfir

2017-11-01

Evaporation from bare soils under laboratory conditions is generally studied using containers of regular shapes where the vertical edges are parallel to the flow lines in the drying domain. The main objective of this study was to investigate the impact of irregular container shapes, for which the flow lines either converge or diverge toward the surface. Evaporation from initially saturated sand and sandy loam soils packed in cones and inverted cones was compared to evaporation from corresponding cylindrical columns. The initial evaporation rate was higher in the cones, and close to potential evaporation. At the end of the experiment, the cumulative evaporation depth in the sand cone was equal to that in the column but higher than in the inverted cone, while in the sandy loam, the order was cone > column > inverted cone. By comparison to the column, stage 1 evaporation was longer in the cones, and practically similar in the inverted cones. Stage 2 evaporation rate decreased with the increase of the evaporating surface area. These results were more pronounced in the sandy loam. For the sand column, the transition between stage 1 and stage 2 evaporation occurred when the depth of the saturation front was approximately equal to the characteristic length of the soil. However, for the cone and the inverted cone, it occurred for a shallower depth of the saturation front. It seems therefore that the concept of the characteristic length derived from the soil hydraulic properties is related to drying systems of regular shapes.

Operational experience of decommissioning techniques for research reactors in the United Kingdom

International Nuclear Information System (INIS)

England, M.R.; McCool, T.M.

2002-01-01

In previous co-ordinated research projects (CRP) conducted by the IAEA no distinction was made between decommissioning activities carried out at nuclear power plants, research reactors or nuclear fuel cycle facilities. As experience was gained and technology advanced it became clear that decommissioning of research reactors had certain specific characteristics which needed a dedicated approach. It was within this context that a CRP on Decommissioning Techniques for Research Reactors was launched and conducted by the IAEA from 1997 to 2001. This paper considers the experience gained from the decommissioning of two research reactors during the course of the CRP namely: (a) the ICI Triga Mk I reactor at Billingham UK which was largely complete by the end of the research project and (b) the Argonaut 100 reactor at the Scottish Universities Research and Reactor centre at East Kilbride in Scotland which is currently is the early stages of dismantling/site operations. It is the intention of this paper with reference to the two case studies outlined above to compare the actual implementation of these works against the original proposals and identify areas that were found to be problematical and/or identify any lessons learnt. (author)

Experiment on continuous operation of the Brazilian IEA-R1 research reactor

International Nuclear Information System (INIS)

Freitas Pintaud, M. de

1994-01-01

In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

The primary circuit of the dragon high temperature reactor experiment

International Nuclear Information System (INIS)

Simon, R.

2005-01-01

The 20 MWth Dragon Reactor Experiment was the first HTGR (High Temperature Gas-cooled Reactor) with coated particle fuel. Its purpose was to test fuel and materials for the High Temperature Reactor programmes pursued in Europe 40 years ago. This paper describes the design and construction of the primary (helium) circuit. It summarizes the main design objectives, lists the performance data and explains the flow paths of the heat removal and helium purification systems. The principal circuit accidents postulated are discussed and the choice of the main construction materials is given. (author)

Application of the Dragon reactor experiment to the safety evaluation of current HTR systems

International Nuclear Information System (INIS)

Ashworth, F.P.O.; Faircloth, R.L.

1976-01-01

An important component of the confidence required for the safety assessment of high-temperature reactors is the experimental proof of phenomena such as fission product release or core corrosion. The most convincing experiments are those carried out in a reactor. This paper outlines the scope of experiments relevant to safety which can be done in the Dragon Reactor Experiment and describes as an example the experimental campaign and the current outcome of the work on validating the predictions of caesium release and migration. (author)

Universal evaporation dynamics of a confined sessile droplet

Science.gov (United States)

Bansal, Lalit; Hatte, Sandeep; Basu, Saptarshi; Chakraborty, Suman

2017-09-01

Droplet evaporation under confinement is ubiquitous to multitude of applications such as microfluidics, surface patterning, and ink-jet printing. However, the rich physics governing the universality in the underlying dynamics remains grossly elusive. Here, we bring out hitherto unexplored universal features of the evaporation dynamics of a sessile droplet entrapped in a 3D confined fluidic environment. We show, through extensive set of experiments and theoretical formulations, that the evaporation timescale for such a droplet can be represented by a unique function of the initial conditions. Moreover, using same theoretical considerations, we are able to trace and universally merge the volume evolution history of the droplets along with evaporation lifetimes, irrespective of the extent of confinement. We also showcase the internal flow transitions caused by spatio-temporal variation of evaporation flux due to confinement. These findings may be of profound importance in designing functionalized droplet evaporation devices for emerging engineering and biomedical applications.

Influence of soil surface structure on simulated infiltration and subsequent evaporation

International Nuclear Information System (INIS)

Verplancke, H.; Hartmann, R.; Boodt, M. de

1983-01-01

A laboratory rainfall and evaporation experiment was conducted to study the effectiveness of the soil surface structure on infiltration and subsequent evaporation. The stability of the surface layer was improved through the application of synthetic additives such as bituminous emulsion and a prepolymer of polyurea (Uresol). The soil column where the soil surface was treated with a bituminous emulsion shows a decrease in depth of wetting owing to the water repellency of that additive, and consequently an increased runoff. However, the application of Uresol to the surface layer improved the infiltration. The main reason for these differences is that in the untreated soils there is a greater clogging of macropores originating from aggregate breakdown under raindrop impact in the top layer. The evaporation experiment started after all columns were wetted to a similar soil-water content and was carried out in a controlled environmental tunnel. Soil-water content profiles were established during evaporation by means of a fully automatic γ-ray scanner. It appears that in both treatments the cumulative evaporation was less than in the untreated soil. This was due to the effect of an aggregated and stabilized surface layer. Under a treated soil surface the evaporation remains constant during the whole experiment. However, under an untreated soil surface different evaporation stages were recorded. From these experiments the impression is gained that the effect of aggregating the soil surface is an increase of the saturated hydraulic conductivity under conditions near saturation. On the other hand, a finely structured layer exhibits a greater hydraulic conductivity during evaporation in the lower soil-water potential range than a coarsely aggregated layer. So it may be concluded that, to obtain the maximum benefit from the available water - optimal water conservation - much attention must be given to the aggregation of the top soil and its stability. (author)

IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

International Nuclear Information System (INIS)

2003-01-01

Description: B and W has performed and analysed a series of physics experiments basically concerned with the technology of heterogeneous reactors moderated and cooled by a variable mixture of heavy and light water. A reactor so moderated is termed Spectral Shift Control Reactor (S SCR). In the practical application of this concept, the moderator mixture is rich in heavy water at the beginning of core life, so a relatively large fraction of the neutrons are epithermal and are absorbed in the fertile material. As fuel is consumed, the moderator is diluted with light water. In this way the neutron spectrum is shifted, thereby increasing the proportion of thermal neutrons and the reactivity of the system. The general objective of the S SCR Basic Physics Program was to study the nuclear properties of rod lattices moderated by D 2 O-H 2 O mixtures. The volume ratio of moderator to non-moderator in all lattices was approximately 1.0, and the fuel was either 4%-enriched UO 2 clad in stainless steel or 93%-enriched UO 2 -ThO 2 (Nth/N 15) pellets clad in aluminum. The D 2 O concentration in the moderator ranged from zero to about 90 mole %. The experimental program includes critical experiments with both types of fuel, exponential experiments at room temperature with both types of fuel, exponential experiments at elevated temperatures with the 4%-enriched UO 2 fuel, and neutron age measurements in ThO 2 lattices. The theoretical program included the development of calculation methods applicable to these systems, and the analysis and correlation of the experimental data. A first report provides the results of critical experiments performed under the Spectral Shift Control Reactor Basic Physics Program. A second report documents experimental results and theoretical interpretation of a series of twenty uniform lattice critical experiments in which the neutron spectrum is varied over a fairly broad range. A third report addresses issues that bear on the problems associated with

Evaporators

DEFF Research Database (Denmark)

Knudsen, Hans Jørgen Høgaard

1996-01-01

Type of evaporators. Regulation. Thermal dimensioning. Determination of pressure loss and heat transfer coefficients.......Type of evaporators. Regulation. Thermal dimensioning. Determination of pressure loss and heat transfer coefficients....

FELIX experiments and computational needs for eddy current analysis of fusion reactors

International Nuclear Information System (INIS)

Turner, L.R.

1984-01-01

In a fusion reactor, changing magnetic fields are closely coupled to the electrically-conducting metal structure. This coupling is particularly pronounced in a tokamak reactor in which magnetic fields are used to confine, stabilize, drive, and heat the plasma. Electromagnetic effects in future fusion reactors will have far-reaching implications in the configuration, operation, and maintenance of the reactors. This paper describes the impact of eddy-current effects on future reactors, the requirements of computer codes for analyzing those effects, and the FELIX experiments which will provide needed data for code validation

Electron beam assisted field evaporation of insulating nanowires/tubes

Energy Technology Data Exchange (ETDEWEB)

Blanchard, N. P., E-mail: nicholas.blanchard@univ-lyon1.fr; Niguès, A.; Choueib, M.; Perisanu, S.; Ayari, A.; Poncharal, P.; Purcell, S. T.; Siria, A.; Vincent, P. [Institut Lumière Matière, UMR5306 Université Lyon 1-CNRS, Université de Lyon, 69622 Villeurbanne Cedex (France)

2015-05-11

We demonstrate field evaporation of insulating materials, specifically BN nanotubes and undoped Si nanowires, assisted by a convergent electron beam. Electron irradiation leads to positive charging at the nano-object's apex and to an important increase of the local electric field thus inducing field evaporation. Experiments performed both in a transmission electron microscope and in a scanning electron microscope are presented. This technique permits the selective evaporation of individual nanowires in complex materials. Electron assisted field evaporation could be an interesting alternative or complementary to laser induced field desorption used in atom probe tomography of insulating materials.

Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

International Nuclear Information System (INIS)

Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

2012-01-01

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

High-burn-up fuels for fast reactors. Past experience and novel applications

International Nuclear Information System (INIS)

Weaver, Kevan D.; Gilleland, John; Whitmer, Charles; Zimmerman, George

2009-01-01

Fast reactors in the U.S. routinely achieved fuel burn-ups of 10%, with some fuel able to reach peak burn-ups of 20%, notably in the Experimental Breeder Reactor II and the Fast Flux Test Facility. Maximum burn-up has historically been constrained by chemical and mechanical interactions between the fuel and its cladding, and to some extent by radiation damage and thermal effects (e.g., radiation-induced creep, thermal creep, and radiation embrittlement) that cause the cladding to weaken. Although fast reactors have used several kinds of fuel - including oxide, metal alloy, carbide, and nitride - the vast majority of experience with fast reactors has been using oxide (including mixed oxide) and metal-alloy fuels based on uranium. Our understanding of high-burn-up operation is also limited by the fact that breeder reactor programs have historically assumed that their fuel would eventually undergo reprocessing; the programs thus have not made high burn-up a top priority. Recently a set of novel designs have emerged for fast reactors that require little initial enrichment and no reprocessing. These reactors exploit a concept known as a traveling wave (sometimes referred to as a breed-and-burn wave, fission wave, or nuclear-burning wave). By breeding and using its own fuel in place as it operates, a traveling-wave reactor can obtain burn-ups that approach 50%, well beyond the current base of knowledge and experience. Our computational work on the physics of traveling-wave reactors shows that they require metal-alloy fuel to provide the margins of reactivity necessary to sustain a breed-and-burn wave. This paper reviews operating experience with high-burn-up fuels and the technical feasibility of moving to a qualitatively new burn-up regime. We discuss our calculations on traveling-wave reactors, including those concerning the possible use of thorium. The challenges associated with high burn-up and fluence in fuels and materials are also discussed. (author)

Reactor dynamics experiment of N.S. Mutsu using pseudo random signal. 1

International Nuclear Information System (INIS)

Hayashi, Koji; Nabeshima, Kunihiko; Shinohara, Yoshikuni; Shimazaki, Junya; Inoue, Kimihiko; Ochiai, Masaaki.

1993-10-01

In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, reactor noise experiments using pseudo random binary sequences (PRBS) have been planned, and a preliminary experiment was performed on March 4, 1991 in the first experimental navigation with the aim of checking the experimental procedures and conditions. The experiments using both reactivity and load disturbances were performed at 70 % of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. From the results obtained, we confirmed that (1) the procedures and experimental conditions determined prior to the experiment were suitable for performing the PRBS experiments, (2) when the PRBS disturbances were applied, the plant state remained quite stable, and (3) the quality of the measured data is adequate for the purpose of dynamics analysis. This paper summarizes the planning and preparation of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

Evaporation of Liquid Droplet in Nano and Micro Scales from Statistical Rate Theory.

Science.gov (United States)

Duan, Fei; He, Bin; Wei, Tao

2015-04-01

The statistical rate theory (SRT) is applied to predict the average evaporation flux of liquid droplet after the approach is validated in the sessile droplet experiments of the water and heavy water. The steady-state experiments show a temperature discontinuity at the evaporating interface. The average evaporation flux is evaluated by individually changing the measurement at a liquid-vapor interface, including the interfacial liquid temperature, the interfacial vapor temperature, the vapor-phase pressure, and the droplet size. The parameter study shows that a higher temperature jump would reduce the average evaporation flux. The average evaporation flux can significantly be influenced by the interfacial liquid temperature and the vapor-phase pressure. The variation can switch the evaporation into condensation. The evaporation flux is found to remain relative constant if the droplet is larger than a micro scale, while the smaller diameters in nano scale can produce a much higher evaporation flux. In addition, a smaller diameter of droplets with the same liquid volume has a larger surface area. It is suggested that the evaporation rate increases dramatically as the droplet shrinks into nano size.

Operating experience and maintenance at the TRIGA Mark II LENA reactor

International Nuclear Information System (INIS)

Cingoli, F.; Meloni, S.; Alloni, L.

1986-01-01

A summary of reactor operation and maintenance in the time period 1982-1986 is presented and discussed. Some problems occurred from instrumentated aluminum cladded elements. Both of them presented damage in the cable tubes and one element showed a protuberance in the cladding. They were replaced with stainless - steel cladded ones. Both elements were sealed up in stainless - steel tubes and put away in wells, 3 meters deep, in the reactor room floor. Some minor problems, correlated to the quite aid instrumentation of the console, are reported. The reactor activity in the last four years was conditioned by the developing of the n - n-bar oscillation NADIR experiment. The thermal column was dismantled and rebuilt in consideration of the Nadir experiment necessities and this job is described in detail. The building containing, the target and the void pipe, presented in 1982 Conference, are now completely operating and the experiment is running. (author)

A review of experiments and results from the transient reactor test (TREAT) facility

International Nuclear Information System (INIS)

Deitrich, L. W.

1998-01-01

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

Decommissioning the Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I)

International Nuclear Information System (INIS)

Harper, J.R.; Garde, R.

1981-11-01

The Los Alamos Molten Plutonium Reactor Experiment (LAMPRE I) was decommissioned at the Los Alamos National Laboratory, Los Alamos, New Mexico, in 1980. The LAMPRE I was a sodium-cooled reactor built to develop plutonium fuels for fast breeder applications. It was retired in the mid-1960s. This report describes the decommissioning procedures, the health physics programs, the waste management, and the costs for the operation

Two-phase reduced gravity experiments for a space reactor design

International Nuclear Information System (INIS)

Antoniak, Z.I.

1986-08-01

Future space missions envision the use of large nuclear reactors utilizing either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed, to coordinate all ongoing and planned reduced gravity flow experiments

Spent-fuel pool thermal hydraulics: The evaporation question

International Nuclear Information System (INIS)

Yilmaz, T.P.; Lai, J.C.

1996-01-01

Many nuclear power plants are currently using dense fuel arrangements that increase the number of spent fuel elements stored in their spent-fuel pools (SFPs). The denser spent-fuel storage results in higher water temperatures, especially when certain event scenarios are analyzed. In some of these event scenarios, it is conservative to maximize the evaporation rate, while in other circumstances it is required to minimize the evaporation rates for conservatism. Evaporation is such a fundamental phenomenon that many branches of engineering developed various equations based on theory and experiments. The evaporation rates predicted by existing equations present a wide range of variation, especially at water temperatures >40 degrees C. Furthermore, a study on which equations provide the highest and lowest evaporation rates has not been done until now. This study explores the sensitivity of existing evaporation equations to various parameters and recommends the limiting evaporation equations for use in the solution of SFP thermal problems. Note that the results of this study may be applicable to a much wider range of applications from irrigation ponds, cooling lakes, and liquid-waste management to calculating adequate air exchange rate for swimming pools and health spas

Development of dynamic simulator for thermosyphon evaporator process with an application

International Nuclear Information System (INIS)

Shimizu, Yoshiaki; Tsutsui, Tenson.

1986-06-01

A dynamic simulator has been developed for radwaste evaporator system in the Research Reactor Institute of Kyoto University. Under mild assumptions, two-phase flow model of the thermosyphon evaporator was shown to be modelled by a set of ordinary and algebraic equations. Through a structure analysis of such equations, a compact but efficient computer program was realized using FORTRAN computer language. By comparing numerical results with experimental ones, reliability of the model has been examined. Furthermore, mentioning several applications imbedded into the developed simulator, a bi-objective optimal problem was formulated generally, and then solved numerically through a practical procedure. It is expected that such a consideration is helpful for the radwaste management in practice. (author)

Elimination of inter-discharge helium glow discharge cleaning with lithium evaporation in NSTX

Directory of Open Access Journals (Sweden)

R. Maingi

2017-08-01

Full Text Available Operation in the National Spherical Torus Experiment (NSTX typically used either periodic boronization and inter-shot helium glow discharge cleaning (HeGDC, or inter-shot lithium evaporation without boronization, and initially with inter-shot HeGDC. To assess the viability of operation without HeGDC, dedicated experiments were conducted in which Li evaporation was used while systematically shrinking the HeGDC between shots from the standard 10min to zero (10→6.5→4→0. Good shot reproducibility without HeGDC was achieved with lithium evaporations of 100mg or higher; evaporations of 200–300mg typically resulted in very low ELM frequency or ELM-free operation, reduced wall fueling, and improved energy confinement. The use of HeGDC before lithium evaporation modestly reduced Dα in the outer scrape-off layer, but not at the strike point. Pedestal electron and ion temperature also improved modestly, suggesting that HeGDC prior to lithium evaporation is a useful tool for experiments that seek to maximize plasma performance.

Burnable poisons in the light water reactor design, microburnup experiments and calculations. Part of a coordinated programme on burnup calculations and experiments for thermal reactors

International Nuclear Information System (INIS)

Penndorf, K.

1976-04-01

Investigations on Research Agreement N 1519/CF (1.8.1974 - 31.7.1975) entitled ''Burnable poisons in light water reactor design, microburnup experiments and calculations'' were carried out in the frame of the IAEA's coordinated research programme on ''Burn-up calculation and experiments for thermal reactors''. The theoretical and experimental work on application of solid burnable poison used for reduction of the amount of boric acid necessary to control of PWR or to lower the number of control rods needed in a BWR. Solid burnable poisons are needed in present PWR designs for the reduction of the boron acid concentration in order to prevent positive coefficients of reactivity. The special operational conditions of a ship reactor lead to the application of this kind of poison for compensation of almost all burnup reactivity. This strengthens the necessity of a very accurate and many dimensional calculations because an appropriate binding of reactivity has to be kept over the whole cycle time. Several burnup experiments had been run in the 15 MW material test reactor FRG-II. The following devices have been irradiated: poison pins within and without PWR fuel pin lattice segments and fuel pins containing pellets with a poison core. Measurements of reactivity, fluence, fission product concentration have been performed. Methods applied were γ-scanning and neutron pulse, radiography and transmission measurement techniques. Evaluation of the experiments was done by one and two dimensional Ssub(N) transport burnup calculations. In parallel a collision probability transport burnup code for current PWR design work is being developed, the main feature of which is economy in manpower and computer time

New Reactor Siting, Licensing and Construction Experience. Proceedings of the 2. CNRA International Workshop on 'New Reactor Siting, Licensing and Construction Experience'

International Nuclear Information System (INIS)

2013-01-01

This report documents the proceedings from the 2. Workshop on New Reactor Siting, Licensing and Construction Experience. A total of 45 specialists from 16 countries and international organisations attended. The meeting was sponsored by the OECD Nuclear Energy Agency Committee on Nuclear Regulatory Activities and hosted by the US Nuclear Regulatory Commission (U.S.NRC). The objectives of the workshop were to provide a forum to exchange information on lessons learned from siting, licensing and constructing new nuclear power plants around the world. Key focus areas included siting practices and regulatory positions that have been enhanced as a result of the Fukushima accident; lessons learned from licensing and design review approaches and challenges, construction experience and recommendations for regulatory oversight; and regulatory cooperation on generic and design specific issues through the MDEP specific working groups. The workshop was structured in 4 technical sessions, each followed by ample time for panel discussions. The first technical session was devoted to regulatory cooperation on generic and design specific issues, MDEP working groups (EPR, AP1000), vendor inspection co-operation, digital I and C, and codes and standards. The second technical session was intended to discuss and share regulatory positions on siting practices and enhancements as a result of lessons learned from Fukushima accident. The third technical session addressed the construction experience and regulatory oversight of new reactor construction activities. And the fourth technical session included presentations on the lessons learned from regulatory licensing reviews of new reactor designs

State system experience with safeguarding power reactors

International Nuclear Information System (INIS)

Roehnsch, W.

1982-01-01

This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

Flash evaporator

OpenAIRE

1997-01-01

A device and method for flash evaporating a reagent includes an evaporation chamber that houses a dome on which evaporation occurs. The dome is solid and of high thermal conductivity and mass, and may be heated to a temperature sufficient to vaporize a specific reagent. The reagent is supplied from an external source to the dome through a nozzle, and may be supplied as a continuous stream, as a shower, and as discrete drops. A carrier gas may be introduced into the evaporation chamber and cre...

Operational experience of the Marcoule reactors; Experience d'exploitation des reacteurs de Marcoule

Energy Technology Data Exchange (ETDEWEB)

Conte, F [Commissariat a l' Energie Atomique, Centre de Production de Plutonium, Marcoule (France). Centre d' Etudes Nucleaires

1963-07-01

The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [French] Les resultats atteints apres trois ans de fonctionnement des reacteurs G-2/G-3 permettent une accumulation considerable de l'experience d'exploitation de ces reacteurs. Les principales originalites: - caisson en beton precontraint - chargement en marche - surveillance automatique des temperatures sont largement justifiees par l'exploitation actuelle. L'auteur confirme l'interet de ces solutions d'avant-garde et en tire des conclusions pour les etudes de futures centrales nucleaires. (auteur)

Quality assurance program plan for the Reactor Research Experiment Programs (RREP)

International Nuclear Information System (INIS)

Pipher, D.G.

1982-05-01

This document describes the Quality Assurance Program plans which will be applied to tasks on Reactor Research Experiments performed on Sandia National Laboratories' reactors. The program provides for individual project or experiment quality plan development and allows for reasonable plan flexibility and maximum plan visibility. Various controls and requirements in this program plan are considered mandatory on all features which are identified as important to public health and safety (Level I). It is the intent of this document that the Quality Assurance program comprise those elements which will provide adequate assurance that all components, equipment, and systems of the experiments will perform as designed, and hence prevent delays and costs due to rejections or failures

Waste Feed Evaporation Physical Properties Modeling

International Nuclear Information System (INIS)

Daniel, W.E.

2003-01-01

This document describes the waste feed evaporator modeling work done in the Waste Feed Evaporation and Physical Properties Modeling test specification and in support of the Hanford River Protection Project (RPP) Waste Treatment Plant (WTP) project. A private database (ZEOLITE) was developed and used in this work in order to include the behavior of aluminosilicates such a NAS-gel in the OLI/ESP simulations, in addition to the development of the mathematical models. Mathematical models were developed that describe certain physical properties in the Hanford RPP-WTP waste feed evaporator process (FEP). In particular, models were developed for the feed stream to the first ultra-filtration step characterizing its heat capacity, thermal conductivity, and viscosity, as well as the density of the evaporator contents. The scope of the task was expanded to include the volume reduction factor across the waste feed evaporator (total evaporator feed volume/evaporator bottoms volume). All the physical properties were modeled as functions of the waste feed composition, temperature, and the high level waste recycle volumetric flow rate relative to that of the waste feed. The goal for the mathematical models was to predict the physical property to predicted simulation value. The simulation model approximating the FEP process used to develop the correlations was relatively complex, and not possible to duplicate within the scope of the bench scale evaporation experiments. Therefore, simulants were made of 13 design points (a subset of the points used in the model fits) using the compositions of the ultra-filtration feed streams as predicted by the simulation model. The chemistry and physical properties of the supernate (the modeled stream) as predicted by the simulation were compared with the analytical results of experimental simulant work as a method of validating the simulation software

Combined potential of future long-baseline and reactor experiments

International Nuclear Information System (INIS)

Huber, P.; Lindner, M.; Rolinec, M.; Schwetz, T.; Winter, W.

2005-01-01

We investigate the determination of neutrino oscillation parameters by experiments within the next ten years. The potential of conventional beam experiments (MINOS, ICARUS, OPERA), superbeam experiments (T2K, NOνA), and reactor experiments (D-CHOOZ) to improve the precision on the 'atmospheric' parameters Δm 31 2 , θ 23 , as well as the sensitivity to θ 13 are discussed. Further, we comment on the possibility to determine the leptonic CP-phase and the neutrino mass hierarchy if θ 13 turns out to be large

IRPhE - International Reactor Physics Experiments database

International Nuclear Information System (INIS)

Sartori, E.

2004-01-01

The OECD/NEA Nuclear Science Committee (NSC) has identified the need to establish international databases containing all the important experiments that are available for sharing among the specialists and has set up or sponsored specific activities to achieve this. The aim is to preserve them in an agreed standard format in computer accessible form, to use them for international activities involving validation of current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques. It is a significant saving results from disseminating a standard benchmark set to be used worldwide. A framework for professionals that use the standard benchmark set to validate and verify modeling codes and data for radiation transport, criticality safety and reactor physics applications guarantees a comparative set of analyses. It represents also a good basis for pinpointing important gaps and where efforts should be concentrated and ensures knowledge and competence preservation, management and transfer in nuclear science and engineering. A large number of experimentalists, physicists, evaluators, modelers have devoted large amounts of their efforts and competencies to produce the data on which the methods we are using today are based. These data are far from having been exploited fully for the different nuclear and radiation technologies. This wealth of information needs to be preserved in a form more easily exploitable by modern information technology and for use in connection with novel and refined computational models with limitations of the past removed. These data will form the basis for the studies of more advanced nuclear technology, will be instrumental in identifying areas where there is a lack of knowledge and thus provide support to justifying new experiments that would reduce design uncertainties and consequently costs. Improvement of

Flashing evaporation under different pressure levels

International Nuclear Information System (INIS)

Liao, Yixiang; Lucas, Dirk; Krepper, Eckhard; Rzehak, Roland

2013-01-01

Highlights: • CFD simulation based on two-fluid model for flashing boiling inside a vertical pipe. • Effect of pressure level on the maximum thermal energy available for evaporation. • Effect of presumed bubble size on the onset of flashing as well as evaporation rate. • Effect of pressure level on the critical bubble size that can start stable flashing. • Effect of pressure level on nucleation rate and mechanism. - Abstract: Flashing evaporation of water inside a vertical pipe under four pressure levels is investigated both experimentally and numerically. In the experiment depressurization is realized through a blow-off valve, and the evaporation rate is controlled by the opening rate and degree of the valve. In the CFD simulation phase change is assumed to be caused by thermal heat transfer between steam–water interface and the surrounding water. Consequently, the evaporation rate is determined by heat transfer coefficient, interfacial area density as well as liquid superheat degree. The simulated temporal course of cross-section averaged steam volume fraction is compared with the measured one. It is found that the increasing rate and maximum value of steam volume fraction is over-predicted under low-pressure conditions, which is mainly caused by the neglect of bubble growth in the mono-dispersed simulation. The agreement is notably improved by performing poly-dispersed simulations with the inhomogeneous MUSIG approach (IMUSIG). On the other hand an underestimation of the maximum steam volume fraction is observed in high-pressure cases, since the contribution of nucleation to the total steam generation rate becomes large as the system pressure increases. Reliable models for nucleation rate as well as bubble detachment size are indispensable for reliable predictions. An effect of the system pressure level on the nucleation mechanism is observed in the experiment

Cooling clothing utilizing water evaporation

DEFF Research Database (Denmark)

Sakoi, Tomonori; Tominaga, Naoto; Melikov, Arsen Krikor

2014-01-01

. To prevent wet discomfort, the T-shirt was made of a polyester material having a water-repellent silicon coating on the inner surface. The chest, front upper arms, and nape of the neck were adopted as the cooling areas of the human body. We conducted human subject experiments in an office with air......We developed cooling clothing that utilizes water evaporation to cool the human body and has a mechanism to control the cooling intensity. Clean water was supplied to the outer surface of the T-shirt of the cooling clothing, and a small fan was used to enhance evaporation on this outer surface...... temperature ranging from 27.4 to 30.7 °C to establish a suitable water supply control method. A water supply control method that prevents water accumulation in the T-shirt and water dribbling was validated; this method is established based on the concept of the water evaporation capacity under the applied...

Maximizing the use of research reactors in training power reactor operating staff with special reference to US experience

International Nuclear Information System (INIS)

Cox, J.A.

1976-01-01

Research reactors have been used in training nuclear power plant personnel for many years. Using the experience in the United States of America a programme is proposed that will maximize the training conducted at a research reactor and lessen the time that the staff must spend training elsewhere. The programme is adaptable to future training of replacement staff and for staff retraining. (author)

Complementary role of critical integral experiment and power reactor start-up experiments for LMFBR neutronics data and method validation

International Nuclear Information System (INIS)

Salvatores, M.

1986-09-01

Both critical experiments and power reactor results play at present a complementary role in reducing the uncertainties in Key design parameters for LMFBR, which can be relevant for the economic performances of this type of reactors

Exploring the correlation between annual precipitation and potential evaporation

Science.gov (United States)

Chen, X.; Buchberger, S. G.

2017-12-01

The interdependence between precipitation and potential evaporation is closely related to the classic Budyko framework. In this study, a systematic investigation of the correlation between precipitation and potential evaporation at the annual time step is conducted at both point scale and watershed scale. The point scale precipitation and potential evaporation data over the period of 1984-2015 are collected from 259 weather stations across the United States. The watershed scale precipitation data of 203 watersheds across the United States are obtained from the Model Parameter Estimation Experiment (MOPEX) dataset from 1983 to 2002; and potential evaporation data of these 203 watersheds in the same period are obtained from a remote-sensing algorithm. The results show that majority of the weather stations (77%) and watersheds (79%) exhibit a statistically significant negative correlation between annual precipitation and annual potential evaporation. The aggregated data cloud of precipitation versus potential evaporation follows a curve based on the combination of the Budyko-type equation and Bouchet's complementary relationship. Our result suggests that annual precipitation and potential evaporation are not independent when both Budyko's hypothesis and Bouchet's hypothesis are valid. Furthermore, we find that the wet surface evaporation, which is controlled primarily by short wave radiation as defined in Bouchet's hypothesis, exhibits less dependence on precipitation than the potential evaporation. As a result, we suggest that wet surface evaporation is a better representation of energy supply than potential evaporation in the Budyko framework.

System Requirements Document for the Molten Salt Reactor Experiment

Energy Technology Data Exchange (ETDEWEB)

Aigner, R.D.

2000-04-01

The purpose of the conversion process is to convert the {sup 233}U fluoride compounds that are being extracted from the Molten Salt Reactor Experiment (MSRE) equipment to a stable oxide for long-term storage at Bldg. 3019.

Spray and evaporation characteristics of ethanol and gasoline direct injection in non-evaporating, transition and flash-boiling conditions

International Nuclear Information System (INIS)

Huang, Yuhan; Huang, Sheng; Huang, Ronghua; Hong, Guang

2016-01-01

Highlights: • Sprays can be considered as non-evaporating when vapour pressure is lower than 30 kPa. • Ethanol direct injection should only be applied in high temperature engine environment. • Gasoline spray collapses at lower fuel temperature (350 K) than ethanol spray does (360 K). • Flash-boiling does not occur when fuel temperature reaches boiling point until ΔT is 14 K. • Not only spray evaporation mode but also breakup mechanism change with fuel temperature. - Abstract: Ethanol direct injection plus gasoline port injection (EDI + GPI) represents a more efficient and flexible way to utilize ethanol fuel in spark ignition engines. To exploit the potentials of EDI, the mixture formation characteristics need to be investigated. In this study, the spray and evaporation characteristics of ethanol and gasoline fuels injected from a multi-hole injector were investigated by high speed Shadowgraphy imaging technique in a constant volume chamber. The experiments covered a wide range of fuel temperature from 275 K (non-evaporating) to 400 K (flash-boiling) which corresponded to cold start and running conditions in an engine. The spray transition process from normal-evaporating to flash-boiling was investigated in greater details than the existed studies. Results showed that ethanol and gasoline sprays demonstrated the same patterns in non-evaporating conditions. The sprays could be considered as non-evaporating when vapour pressure was lower than 30 kPa. Ethanol evaporated more slowly than gasoline did in low temperature environment, but they reached the similar evaporation rates when temperature was higher than 375 K. This suggested that EDI should only be applied in high temperature engine environment. For both ethanol and gasoline sprays, when the excess temperature was smaller than 4 K, the sprays behaved the same as the subcooled sprays did. The sprays collapsed when the excess temperature was 9 K. Flash-boiling did not occur until the excess temperature

Evaporation of tungsten in vacuum at low hydrogen and water vapor pressures

International Nuclear Information System (INIS)

Andrievskij, R.A.; Galkin, E.A.; Khromonozhkin, V.V.

1981-01-01

The results of experimental investigations of tungsten evaporation rates in the temperature range 1650-2500 K, partial hydrogen and water vapours pressures 1x10 -5 -10 Pa are presented. Experi-- mental plant, equipment employed and radiometric technique of tungsten evaporation study are described. The dependences of evaporation rate and probabilities of tungsten oxidation by residual vacuum water vapours and dependences of tungsten evaporation rate on partial hydrogen and water vapours pressures are determined [ru

French experience concerning expansion compensating devices on the primary systems of nuclear reactors

International Nuclear Information System (INIS)

Vrillon, B.; Raynal, A.

1980-01-01

French experience in the use of large expansion bellows in the presence of hot sodium is extremely limited. This stems from the 'pool' structure of the primary circuit, adopted in France to eliminate the need to solve expansion problems affecting the primary piping of loop reactors. Furthermore, until the present time, the use of bellows on secondary circuits has neither been implemented nor considered. A few bellows nevertheless exist on the Phenix and Super-Phenix reactors, and these perform separation functions, for example, between sodium at different temperature and/or pressures, or tightness functions in gaseous environment at the component penetrations in the slabs. The dimension criteria applied to these bellows are the general rules for structural dimensioning. Since they do not form part of a circuit wall, they do not need to be discussed. Note, however, that these components have not raised any particular problems thus far. Expansion bellows exist in France on the primary circuits of certain nuclear reactors of the natural uranium/graphite/gas type. These reactors have been in operation for many years, and some lessons can be drawn from this experience in the use of bellows in representative conditions on power reactor circuits. Liquid sodium raises specific problems with respect to circuit operation and material behavior. However, many problems in the use of bellows are independent of the fluid conveyed in the circuits. This is why the experience gained with gas type power reactors appears to be useful in considering the possible future use of bellows on sodium reactor circuits

Plant experience of experimental fast reactor 'Joyo'

International Nuclear Information System (INIS)

1982-01-01

The experimental fast reactor ''JOYO'' installed in Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan completed its operation using the first core (called MK-I core) in December, 1981, and the works to transfer to MK-2 core have been performed since January, 1982. In this report, the experiences obtained through the construction, test and operation of ''JOYO'' over 12 years from the start of erection in 1970 to the termination of operation in 1981 are described. The contents of the report are divided into design, construction, the outline of facilities, testing, operating and maintenance experiences, and the topics on MK-I operation. As for the construction, the design changes performed before the start of manufacture or construction and the improvement and trouble restoring works implemented at the start of overall functional tests are reported. As for testing, overall functional tests, criticality test, low power test and power increasing test are described in detail. The number of test items of overall functional testing reached 266. The rated output operation of the reactor at 75 MW was performed six times in 1980 and 1981 until the termination of operation. No fuel failure was detected in MK-I operation, and the stable operation performance of the FBR was proved through MK-I operation. The topics on the MK-I operation includes natural circulation test, the measurement of total leakage rate for the containment vessel, and wear-marks which are the trace of wear due to the contact of fuel pins with the wires wound around the adjacent fuel pins, found in the post irradiation examination of fuel. (Wakatsuki, Y.)

French gas cooled reactor experience with moisture ingress

International Nuclear Information System (INIS)

Bastien, D.; Brie, M.

1995-01-01

During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

Experience with mechanical segmentation of reactor internals

International Nuclear Information System (INIS)

Carlson, R.; Hedin, G.

2003-01-01

Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

Experiences with fast breeder reactor education in laboratory and short course settings

International Nuclear Information System (INIS)

Waltar, A.E.

1983-01-01

The breeder reactor industry throughout the world has grown impressively over the last two decades. Despite the uncertainties in some national programs, breeder reactor technology is well established on a global scale. Given the magnitude of this technological undertaking, there has been surprisingly little emphasis on general breeder reactor education - either at the university or laboratory level. Many universities assume the topic too specialized for including appropriate courses in their curriculum - thus leaving students entering the breeder reactor industry to learn almost exclusively from on-the-job experience. The evaluation of four course presentations utilizing visual aids is presented

The SM and MIR reactors operation experience

International Nuclear Information System (INIS)

Kuprienko, V.A.; Klinov, A.V.; Svyatkin, M.N.; Shamardin, V.K.

1995-01-01

The SM and MIR operation experience show that continuous work on the problem of ageing, in all its aspects, allows for prolongation of the research plant life cycle by several folds as compared to the initial project. The redesigned SM-3 reactor will operate for another 20 years. The similar result is expected from the MIR planned reconstruction which scope will be the topic of future presentations. (orig.)

Further experience in simulation of rod drop experiments in the Loviisa and Mochovce reactors

International Nuclear Information System (INIS)

Siltanen, P.; Kaloinen, E.; Tanskanen, A.; Mattila, R.

2001-01-01

Simulations of reactor scram experiments using the 3-dimensional kinetics code HEXTRAN have been updated for the initial cores of Loviisa-1 and 2 Mochovce-1 and have been extended to burned cores of Loviisa-1. In these simulations, the entire experiment is simulated dynamically, including the behaviour of the core, the signal of the ionization chamber, and the inverse point kinetics of the reactivity meter. The predicted output of the reactivity meter is compared with the output observed during the experiment (Authors)

Test on the reactor with the portable digital reactivity meter for physical experiment

International Nuclear Information System (INIS)

Huang Liyuan

2010-01-01

Test must be performed on the zero power reactor During the development of portable digital reactivity meter for physical experiment, in order to check its measurement function and accuracy. It describes the test facility, test core, test methods, test items and test results. The test results show that the instrument satisfy the requirements of technical specification, and satisfy the reactivity measurement in the physical experiments on reactors. (authors)

Description of the french graphite reactor and of the experiments performed in 1956

International Nuclear Information System (INIS)

Bussac, J.; Leduc, C.; Zaleski, C.P.

1957-01-01

This paper is an introduction to the experiments performed on the G1 reactor, experiments fully described in the papers following (670 'B to P'). The main results are given together with some comments. The neutronic parameters of the core, a description of the most important structures, and a few words of the tests leading to normal operation of the reactor under load complete our survey. (author) [fr

Gas cooled reactors

International Nuclear Information System (INIS)

Kojima, Masayuki.

1985-01-01

Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

Latest Results from the Daya Bay Reactor Neutrino Experiment

CERN Multimedia

CERN. Geneva

2014-01-01

Among all the fundamental particles that have been experimentally observed, neutrinos remain one of the least understood. The Daya Bay Reactor Neutrino Experiment in China consists of eight identical detectors placed underground at different baselines from three groups of nuclear reactors, a configuration that is ideally suited for studying the properties of these elusive particles. This talk will present three sets of results that have just recently been released by the Daya Bay Collaboration: (i) a precision measurement of the oscillation parameters that drive the disappearance of electron antineutrinos at short baselines, (ii) a search for sterile neutrino mixing, and (iii) a high-statistics determination of the absolute flux and spectrum of reactor-produced electron antineutrinos. All of these results extend the limits of our knowledge in their respective areas and thus shed new light on neutrinos and the physics that surround them.

Evaporation behaviour of different organic effluents from open surfaces.

Science.gov (United States)

Jhorar, B S; Malik, R S

1993-01-01

Production of large quantities of effluents from different industrial units and the problems of their disposal necessitated this evaporation study. The evaporation of water, sewage water, oil refinery effluent, papermill effluent and liquor distillery effluent was observed in glass beakers when placed (i) in an oven at 60 degrees C and (ii) in screen house for 30 days, by periodically weighing of the beakers. In other experiments, the effect of increasing the frequency of stirring on increasing the evaporation efficiency of the liquor distillery effluent (ELD) was examined in detail. All of the organic effluents except ELD had similar evaporation behaviours as water, but formation of a self-forming film caused the evaporation of ELD to be considerably lower. Resistance to evaporation caused by this film was found to be a decreasing function of the frequency of stirring. This study has a bearing on improving the efficiency of evaporation lagoons, and three stirrings in a day with a manually drawn stirrer in a full-scale lagoon are proposed as a practical and economically viable technique to save 44% of lagoon land in arid and semi-arid regions of the world.

Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mueller, Don [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-12-01

In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 ⁷LiF + BeF₂) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE. The key finding of this work is that, for both solid and liquid fueled fluoride salt reactors, radiative capture in ⁷Li is the most significant contributor to potential bias in neutronics calculations within the FLiBe salt.

Investigation of safety measures to severe accident of Fast Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

So as to plan the accident management to severe accident of Fast Breeder Reactor (FBR), it is primary important to understand the progression of severe accident (SA) precisely. In this study, it has been aimed to reveal two items that work as keys in the evaluation of SA in sodium cooled FBR. One is the cool-ability of degraded core on the core support plate by sodium natural circulation in the post accident heat removal (PAHR) phase. An obstacle that hinders the smooth heat transfer from fuel debris to coolant is the formation of sodium-uranate by chemical reaction between sodium and fuel. Following the measurement of physical values of sodium-uranate in FY 2011, experiments has been performed to reveal the conditions for sodium-uranate formation on fuel debris in sodium pool simulating the actual situation of the degraded core. The cool-ability of the debris bed was analyzed using the Lipinski 1-D model. Another research performed in this study is the measurement of fission product (cesium and antimony) evaporation rates from FBR fuel as a function of temperature, because presently the fission product evaporation rates data for LWR is also temporarily used for FBR SA analysis. The measurement was performed using the irradiated fuels in the Test Reactor JOYO. (author)

Simulation of Diffusive Lithium Evaporation Onto the NSTX Vessel Walls

International Nuclear Information System (INIS)

Stotler, D.P.; Skinner, C.H.; Blanchard, W.R.; Krstic, P.S.; Kugel, H.W.; Schneider, H.; Zakharov, L.E.

2010-01-01

A model for simulating the diffusive evaporation of lithium into a helium filled NSTX vacuum vessel is described and validated against an initial set of deposition experiments. The DEGAS 2 based model consists of a three-dimensional representation of the vacuum vessel, the elastic scattering process, and a kinetic description of the evaporated atoms. Additional assumptions are required to account for deuterium out-gassing during the validation experiments. The model agrees with the data over a range of pressures to within the estimated uncertainties. Suggestions are made for more discriminating experiments that will lead to an improved model.

Evaluation of performance of select fusion experiments and projected reactors. Final report

International Nuclear Information System (INIS)

Miley, G.H.

1978-10-01

The performance of NASA Lewis fusion experiments (SUMMA and Bumpy Torus) is compared with other experiments and that necessary for a power reactor. Key parameters cited are gain (fusion power/input power) and the time average fusion power, both of which may be more significant for real fusion reactors than the commonly used Lawson parameter. The NASA devices are over 10 orders of magnitude below the required powerplant values in both gain and time average power. The best experiments elsewhere are also as much as 4 to 5 orders of magnitude low. However, the NASA experiments compare favorably with other alternate approaches that have received less funding than the mainline experiments. The steady-state character and efficiency of plasma heating are strong advantages of the NASA approach. The problem, though, is to move ahead to experiments of sufficient size to advance in gain and average power parameters

Evaporational losses under different soil moisture regimes and atmospheric evaporativities using tritium

International Nuclear Information System (INIS)

Saxena, P.; Chaudhary, T.N.; Mookerji, P.

1991-01-01

Tritium as tracer was used in a laboratory study to estimate the contribution of moisture from different soil depths towards actual soil water evaporation. Results indicated that for comparable amounts of free water evaporation (5 cm), contribution of moisture from 70-80 cm soil layer towards total soil moisture loss through evaporation increased nearly 1.5 to 3 folds for soils with water table at 90 cm than without water table. Identical initial soil moistures were exposed to different atmospheric evaporativities. Similarly, for a given initial soil moisture status, upward movement of moisture from 70-80 cm soil layer under low evaporativity was nearly 8 to 12 times that of under high evaporativity at 5 cm free water evaporation value. (author). 6 refs., 4 tabs., 2 figs

Reactors Project Delivery: The Value of Experiance

International Nuclear Information System (INIS)

Stosic, V. Zoran

2014-01-01

State of Affairs: Energy Potential and Density versus Environmental Load of different Energy Sources, Development of Fuel into Energy/Electricity Generation, Production Costs of Electricity, Contributions of Nuclear Energy to Security of Energy Supply, Recent Nuclear Development, Public Support growing again. Projects Status: Reactors under Construction, Different Projects Industrial Schemes, Projects Overview. The Value of Experience: Licensing, Standardization on Early Engineering Activities, Supply Chain and Manufacturing of Heavy Components, Installation, Procurement. (author)

Neutron irradiation experiments for fusion reactor materials through JUPITER program

International Nuclear Information System (INIS)

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

1998-01-01

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

Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

International Nuclear Information System (INIS)

1975-12-01

The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

Heat and mass transfer analogies for evaporation models at high evaporation rate

OpenAIRE

Trontin , P.; Villedieu , P.

2014-01-01

International audience; In the framework of anti and deicing applications, heated liquid films can appear above the ice thickness, or directly above the wall. Then, evaporation plays a major role in the Messinger balance and evaporated mass has to be predicted accurately. Unfortunately, it appears that existing models under-estimate evaporation at high temperature. In this study, different evaporation models at high evaporation rates are studied. The different hypothesis on which these models...

Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

International Nuclear Information System (INIS)

Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

1981-01-01

This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

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.

An experimental study for the interface shear stress of near vertical air-water separated flow on evaporation

International Nuclear Information System (INIS)

Kwon, H.; Park, G. C.

2000-01-01

The object of experiment is improved model of evaporative heat transfer coefficient using interfacial friction factor on evaporation. Experiments have been conducted with near-vertical(87 .deg.) flat plate on evaporation for air-water countercurrent stratified flow. Experiment facility is consisted of 1.7m length and 0.2 X 0.005m cross section, the one side direct heating system which have 10kw power capacity. The interfacial shear stress, pressure drop and temperatures in test section were measured. These parameters were measured by DP-103 pressure transducer, K-type thermocouple, RTD and Hot Wire Anemometer(HWA). Experimental results were inclination as increased interfacial shear stress with increased the evaporation rate. Interfacial shear stress was increased as increased water flow rate and air flow rate too. For the evaluation of the measured evaporative heat transfer coefficients and physical understanding of the evaporation phenomena, the evaporative heat transfer coefficients were obtained through the simple calculation process by the use of mass transfer coefficient correlation and the experimental data of wavy film surface effect on shear and on evaporation

Packaging and shipment of U.S. breeder reactor experiments

International Nuclear Information System (INIS)

Berger, J.D.

1980-01-01

Irradiation testing of fuels and materials in the Fast Test Reactor (FTR) required development of a shipping cask (designated T-3) and associated hardware for loading and shipping of these experiments to postirradiation examination facilities. The T-3 shipping-cask program included design, fabrication, and testing of internal cask packages to protect the experiments during loading, shipping, and unloading. The cask was designed for loading in both the vertical and horizontal attitudes

Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

Energy Technology Data Exchange (ETDEWEB)

Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

1968-04-15

Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

Operating experience feedback report: Experience with pump seals installed in reactor coolant pumps manufactured by Byron Jackson

International Nuclear Information System (INIS)

Bell, L.G.; O'Reilly, P.D.

1992-09-01

This report examines the reactor coolant pump (RCP) seal operating experience through August 1990 at plants with Byron Jackson (B-J) RCPs. ne operating experience examined in this analysis included a review of the practice of continuing operation with a degraded seal. Plants with B-J RCPs that have had relatively good experience with their RCP seals attribute this success to a combination of different factors, including: enhanced seal QA efforts, modified/new seal designs, improved maintenance procedures and training, attention to detail, improved seal operating procedures, knowledgeable personnel involved in seal maintenance and operation, reduction in frequency of transients that stress the seals, seal handling and installation equipment designed to the appropriate precision, and maintenance of a clean seal cooling water system. As more plants have implemented corrective measures such as these, the number of B-J RCP seal failures experienced has tended to decrease. This study included a review of the practice of continued operation with a degraded seal in the case of PWR plants with Byron Jackson reactor coolant pumps. Specific factors were identified which should be addressed in order to safety manage operation of a reactor coolant pump with indications of a degrading seal

The evaporative fraction as a measure of surface energy partitioning

Energy Technology Data Exchange (ETDEWEB)

Nichols, W.E. [Pacific Northwest Lab., Richland, WA (United States); Cuenca, R.H. [Oregon State Univ., Corvallis, OR (United States)

1990-12-31

The evaporative fraction is a ratio that expresses the proportion of turbulent flux energy over land surfaces devoted to evaporation and transpiration (evapotranspiration). It has been used to characterize the energy partition over land surfaces and has potential for inferring daily energy balance information based on mid-day remote sensing measurements. The HAPEX-MOBILHY program`s SAMER system provided surface energy balance data over a range of agricultural crops and soil types. The databases from this large-scale field experiment was analyzed for the purpose of studying the behavior and daylight stability of the evaporative fraction in both ideal and general meteorological conditions. Strong linear relations were found to exist between the mid-day evaporative fraction and the daylight mean evaporative fraction. Statistical tests however rejected the hypothesis that the two quantities were equal. The relations between the evaporative fraction and the surface soil moisture as well as soil moisture in the complete vegetation root zone were also explored.

The evaporative fraction as a measure of surface energy partitioning

Energy Technology Data Exchange (ETDEWEB)

Nichols, W.E. (Pacific Northwest Lab., Richland, WA (United States)); Cuenca, R.H. (Oregon State Univ., Corvallis, OR (United States))

1990-01-01

The evaporative fraction is a ratio that expresses the proportion of turbulent flux energy over land surfaces devoted to evaporation and transpiration (evapotranspiration). It has been used to characterize the energy partition over land surfaces and has potential for inferring daily energy balance information based on mid-day remote sensing measurements. The HAPEX-MOBILHY program's SAMER system provided surface energy balance data over a range of agricultural crops and soil types. The databases from this large-scale field experiment was analyzed for the purpose of studying the behavior and daylight stability of the evaporative fraction in both ideal and general meteorological conditions. Strong linear relations were found to exist between the mid-day evaporative fraction and the daylight mean evaporative fraction. Statistical tests however rejected the hypothesis that the two quantities were equal. The relations between the evaporative fraction and the surface soil moisture as well as soil moisture in the complete vegetation root zone were also explored.

Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

2004-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

Modeling, simulation, and analysis of a reactor system for the generation of white liquor of a pulp and paper industry

Directory of Open Access Journals (Sweden)

Ricardo Andreola

2011-02-01

Full Text Available An industrial system for the production of white liquor of a pulp and paper industry, Klabin Paraná Papéis, formed by ten reactors was modeled, simulated, and analyzed. The developed model considered possible water losses by the evaporation and reaction, in addition to variations in the volumetric flow of lime mud across the reactors due to the composition variations. The model predictions agreed well with the process measurements at the plant and the results showed that the slaking reaction was nearly complete at the third causticizing reactor, while causticizing ends by the seventh reactor. Water loss due to slaking reaction and evaporation occurred more pronouncedly in the slaker reactor than in the final causticizing reactors; nevertheless, the lime mud flow remained nearly constant across the reactors.

Using evaporation to control capillary instabilities in micro-systems.

Science.gov (United States)

Ledesma-Aguilar, Rodrigo; Laghezza, Gianluca; Yeomans, Julia M; Vella, Dominic

2017-12-06

The instabilities of fluid interfaces represent both a limitation and an opportunity for the fabrication of small-scale devices. Just as non-uniform capillary pressures can destroy micro-electrical mechanical systems (MEMS), so they can guide the assembly of novel solid and fluid structures. In many such applications the interface appears during an evaporation process and is therefore only present temporarily. It is commonly assumed that this evaporation simply guides the interface through a sequence of equilibrium configurations, and that the rate of evaporation only sets the timescale of this sequence. Here, we use Lattice-Boltzmann simulations and a theoretical analysis to show that, in fact, the rate of evaporation can be a factor in determining the onset and form of dynamical capillary instabilities. Our results shed light on the role of evaporation in previous experiments, and open the possibility of exploiting diffusive mass transfer to directly control capillary flows in MEMS applications.

Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

International Nuclear Information System (INIS)

Naff, S.A.; Schwarz, W.F.

1978-01-01

Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

Evaporator Cleaning Studies

International Nuclear Information System (INIS)

Wilmarth, W.R.

1999-01-01

Operation of the 242-16H High Level Waste Evaporator proves crucial to liquid waste management in the H-Area Tank Farm. Recent operational history of the Evaporator showed significant solid formation in secondary lines and in the evaporator pot. Additional samples remain necessary to ensure material identity in the evaporator pot. Analysis of these future samples will provide actinide partitioning information and dissolution characteristics of the solid material from the pot to ensure safe chemical cleaning

Removal of Sulfate Ion From AN-107 by Evaporation

International Nuclear Information System (INIS)

GJ Lumetta; GS Klinger; DE Kurath; RL Sell; LP Darnell; LR Greenwood; CZ Soderquist; MJ Steele; MW Urie; JJ Wagner

2000-01-01

Hanford low-activity waste solutions contain sulfate, which can cause accelerated corrosion of the vitrification melter and unacceptable operating conditions. A method is needed to selectively separate sulfate from the waste. An experiment was conducted to evaluate evaporation for removing sulfate ion from Tank AN-107 low-activity waste. Two evaporation steps were performed. In the first step, the volume was reduced by 55% while in the second step, the liquid volume was reduced another 22%. Analysis of the solids precipitated during these evaporations revealed that large amounts of sodium nitrate and nitrite co-precipitated with sodium sulfate. Many other waste components precipitated as well. It can be concluded that sulfate removal by precipitation is not selective, and thus, evaporation is not a viable option for removing sulfate from the AN-107 liquid

Study on low pressure evaporation of fresh water generation system model

Energy Technology Data Exchange (ETDEWEB)

Chung, Han Shik; Wibowo, Supriyanto; Shin, Yong Han; Jeong, Hyo Min [Gyeongsang National University, Tongyeong (Korea, Republic of); Fajar, Berkah [University of Diponegoro, Semarang (Indonesia)

2012-02-15

A low pressure evaporation fresh water generation system is designed for converting brackish water or seawater into fresh water by distillation in low pressure and temperature. Distillation through evaporation of feed water and subsequent vapor condensation as evaporation produced fresh water were studied; tap water was employed as feed water. The system uses the ejector as a vacuum creator of the evaporator, which is one of the most important parts in the distillation process. Hence liquid can be evaporated at a lower temperature than at normal or atmospheric conditions. Various operating conditions, i.e. temperature of feed water and different orifice diameters, were applied in the experiment to investigate the characteristics of the system. It was found that these parameters have a significant effect on the performance of fresh water generation systems with low pressure evaporation.

Mixed phase evaporation source

International Nuclear Information System (INIS)

1975-01-01

Apparatus for reducing convection current heat loss in electron beam evaporator is described. A material to be evaporated (evaporant) is placed in the crucible of an electron beam evaporation source along with a porous mass formed of a powdered or finely divided solid to act as an impedance to convection currents. A feed system is employed to replenish the supply of evaporant as it is vaporized

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)

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

International Nuclear Information System (INIS)

2017-01-01

The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

Energy Technology Data Exchange (ETDEWEB)

Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

1974-07-01

Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

Experimental study of falling film evaporation in large scale rectangular channel

International Nuclear Information System (INIS)

Huang, X.G.; Yang, Y.H.; Hu, P.

2015-01-01

Highlights: • This paper studies the falling film evaporation in large scale rectangular channel experimentally. • The effects of air flow rate, film temperature and film flow rate on falling film evaporation are analyzed. • Increasing the air flow rate is considered as an efficient method to enhance the evaporation rate. • A correlation including the wave effect for falling film evaporation is derived based on heat and mass transfer analogy. - Abstract: The falling film evaporation in a large scale rectangular channel is experimentally studied in this paper for the design and improvement of passive containment cooling system. The evaporation mass transfer coefficient h D is obtained by the evaporation rate and vapor partial pressure difference of film surface and air bulk. The experimental results indicate that increasing of air flow rate appears to enhance h D , while the film temperature and film flow rate have little effect on h D . Since the wave effect on evaporation is noticed in experiment, the evaporation mass transfer correlation including the wave effect is developed on the basis of heat and mass transfer analogy and experimental data

Further analysis of the zero-energy experiment on the Dragon reactor

International Nuclear Information System (INIS)

Woloch, F.; Neuberger, W.

1978-01-01

The analysis of the Zero-Energy Experiments performed on the Dragon reactor, a high-temperature reactor of the Organization for Economic Cooperation and Development, has been continued. The first analysis established the main route of calculations within the WIMS-E scheme and was reported elsewhere. This Note presents further calculations showing the merits of a refinement in the number of neutron energy groups, of the use of different condensation spectra, and of transport calculations

Experiments in power distribution control on the IRT-2000 reactor

International Nuclear Information System (INIS)

Filipchuk, E.V.; Potapenko, P.T.; Trofimov, A.P.; Kosilov, A.N.; Neboyan, V.T.; Timokhin, E.S.

1975-01-01

The results from the experimental investigations of a system for regulating the neutron field on a research reactor IRT-2000 are shown. The right of such experiments on a reactor with a little active zone is substantiated. A successful attempt was made in this work to apply primary elements of straight charging in the neutron field regulating system. A system with independent instrumentally local regulators, a system with hard cross connections and a structure with a ''floating'' installation are studied. Serial common industrial regulators BRT-2 were used

Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

Energy Technology Data Exchange (ETDEWEB)

Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

1995-09-01

There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

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.

Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

Energy Technology Data Exchange (ETDEWEB)

Wood, RT

2004-09-27

This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

Aggregation of evaporative fraction by remote sensing from micro to macro scale

NARCIS (Netherlands)

Bastiaanssen, W.G.M.; Pelgrum, H.; Wal, van der T.; Roebeling, R.A.

1996-01-01

The evaporative fraction of the surface energy balance has been favoured as a tool to describe the energy partitioning during daytime. It is shown that the evaporative fraction behaves temporally stable under heterogeneous terrain conditions in the Echival Field Experiment in

Sea water desalination utilizing waste heat by low temperature evaporation

International Nuclear Information System (INIS)

Raha, A.; Srivastava, A.; Rao, I.S.; Majumdar, M.; Srivastava, V.K.; Tewari, P.K.

2007-01-01

Economics of a process is controlled by management of energy and resources. Fresh water has become most valued resource in industries. Desalination is a process by which fresh water resource is generated from sea water or brackish water, but it is an energy intensive process. The energy cost contributes around 25-40% to the total cost of the desalted water. Utilization of waste heat from industrial streams is one of the ecofriendly ways to produce low cost desalted water. Keeping this in mind Low Temperature Evaporation (LTE) desalination technology utilizing low quality waste heat in the form of hot water (as low as 50 deg C) or low pressure steam (0.13 bar) has been developed for offshore and land based applications to produce high purity water (conductivity < 2μS/cm) from sea water. The probability of the scale formation is practically eliminated by operating it at low temperature and controlling the brine concentration. It also does not require elaborate chemical pretreatment of sea water except chlorination, so it has no environmental impact. LTE technology has found major applications in nuclear reactors where large quantity of low quality waste heat is available to produce high quality desalted water for make up water requirement replacing conventional ion exchange process. Successful continuous operation of 30 Te/day LTE desalination plant utilizing waste heat from nuclear research reactor has demonstrated the safety, reliability, extreme plant availability and economics of nuclear desalination by LTE technology. It is also proposed to utilize waste heat from Main Heat Transport (MHT) purification circuit of Advanced Heavy Water Reactor (AHWR) to produce about 250 Te/ day high quality desalinated water by Low Temperature Evaporation (LTE) process for the reactor make up and plant utilization. Recently we have commissioned a 50 Te/day 2-effect low temperature desalination plant with cooling tower where the specific energy and cooling water requirement are

Experiment calculated ascertainment of factors affecting the energy release in IGR reactor core

International Nuclear Information System (INIS)

Kurpesheva, A.M.; Zhotabayev, Zh.R.

2006-01-01

Full text: At present energy supply resources problem is important. Nuclear reactors can, of course, solve this problem, but at the same time there is another issue, concerning safety exploitation of nuclear reactors. That is why, for the last seven years, such experiments as 'Investigation of the processes, conducting severe accidents with core melting' are being carried out at our IGR (impulse graphite reactor) reactor. Leaving out other difficulties of such experiments, it is necessary to notice, that such experiments require more accurate IGR core energy release calculations. The final aim of the present research is verification and correction of the existing method or creation of new method of IGR core energy release calculation. IGR reactor is unique and there is no the same reactor in the world. Therefore, application of the other research reactor methods here is quite useful. This work is based on evaluation of factors affecting core energy release (physical weight of experimental device, different configuration of reactor core, i.e. location of absorbers, initial temperature of core, etc), as well as interference of absorbers group. As it is known, energy release is a value of integral reactor power. During experiments with rays, Reactor power depends on currents of ion production chambers (IPC), located round the core. It is worth to notice that each ion production chamber (IPC) in the same start-up has its own ratio coefficient between IPC current and reactor present power. This task is complicated due to 'IPC current - reactor power' ratio coefficients, that change continuously, probably, because of new loading of experimental facility and different position of control rods. That is why, in order to try about reactor power, before every start-up, we have to re-determine the 'IPC current - reactor power' ratio coefficients for each ion production chamber (IPC). Therefore, the present work will investigate the behavior of ratio coefficient within the

Light water reactor mixed-oxide fuel irradiation experiment

International Nuclear Information System (INIS)

Hodge, S.A.; Cowell, B.S.; Chang, G.S.; Ryskamp, J.M.

1998-01-01

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding

Evaporation Kinetics of Laboratory Generated Secondary Organic Aerosols at Elevated Relative Humidity

Energy Technology Data Exchange (ETDEWEB)

Wilson, Jacqueline M.; Imre, D.; Beranek, Josef; Shrivastava, ManishKumar B.; Zelenyuk, Alla

2015-01-06

Secondary organic aerosols (SOA) dominate atmospheric organic aerosols that affect climate, air quality, and health. Recent studies indicate that, contrary to previously held assumptions, at low relative humidity (RH) these particles are semi-solid and evaporate orders of magnitude slower than expected. Elevated relative humidity has the potential to affect significantly formation, properties, and atmospheric evolution of SOA particles. Here we present a study of the effect of RH on the room-temperature evaporation kinetics of SOA particles formed by ozonolysis of α-pinene and limonene. Experiments were carried out on SOA particles generated, evaporated, and aged at 0%, 50% and 90% RH. We find that in all cases evaporation begins with a relatively fast phase, during which 30% to 70% of the particle mass evaporates in 2 hours, followed by a much slower evaporation rate. Evaporation kinetics at 0% and 50% RH are nearly the same, while at 90% RH a slightly larger fraction evaporates. In all cases, aging the particles prior to inducing evaporation reduces the evaporative losses, with aging at elevated RH leading to more significant effect. In all cases, SOA evaporation is nearly size-independent, providing direct evidence that oligomers play a crucial role in determining the evaporation kinetics.

Heavy metal evaporation kinetics in thermal waste treatment processes

Energy Technology Data Exchange (ETDEWEB)

Ludwig, Ch; Stucki, S; Schuler, A J [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1999-08-01

To investigate the evaporation kinetics of heavy metals, experiments were performed by conventional thermogravimetry and a new method using Inductively Coupled Plasma Optical Emission Spectroscopy (ICP-OES). The new method allows online measurements in time intervals that are typically below one minute. The evaporation of Cd, Cu, Pb, and Zn from synthetic mixtures and filter ashes from municipal solid waste incineration (MSWI) was of major interest. (author) 2 figs., 4 refs.

Factors in the selection of broiler tube materials for a civil fast reactor

Energy Technology Data Exchange (ETDEWEB)

Tyzack, C; Chitty, A

1975-07-01

This paper briefly considers some of the factors which must be balanced in the selection of a boiler tube material for a Civil Fast Reactor. The merits and possible demerits of low alloy ferritic steels and the austenitic Alloy 800 are compared with respect to waterside corrosion resistance, mechanical properties, fabrication and weldability and possible effects of exposure to the sodium environment under normal and fault conditions. It is pointed out that although there is operational experience of most of the materials in boiler superheater applications there is little or none in evaporative regimes. (author)

Advective-diffusive transport of D2O in unsaturated media under evaporation condition

International Nuclear Information System (INIS)

Koarashi, Jun; Atarashi-Andoh, Mariko; Amano, Hikaru; Yamazawa, Hiromi; Iida, Takao

2003-01-01

Advective-diffusive transport of HTO in unsaturated media was investigated empirically using deuterated water (D 2 O) and columns filled with glass beads. The tortuosity factor was evaluated by numerical model calculations corresponding to first experiment for diffusion under no-evaporation condition. Temporal variations in depth profiles of D 2 O concentrations in the columns were observed by second experiment, which considers the transferring and spreading of D 2 O by pore-water flow caused by evaporation. Measurements and model calculations indicated that diffusion was about two times more efficient than dispersion for D 2 O spreading process under this evaporation condition. (author)

Complementarity of integral and differential experiments for reactor physics purposes

International Nuclear Information System (INIS)

Tellier, Henry.

1981-04-01

In this paper, the following topics are studied: uranium 238 effective integral; thermal range uranium 238 capture cross section; Americium 242 m capture cross section. The mentioned examples show that differential and integral experiments are both useful to the reactor physicists

Operating experience feedback report: Progress in scram reduction: Commercial power reactors

International Nuclear Information System (INIS)

Bell, L.G.; O'Reilly, P.D.

1989-03-01

This report documents the results of a trends and patterns analysis of unplanned reactor scrams at commercial US nuclear power reactors from January 1, 1984 to January 1, 1988. Major objectives of this report prepared by the Nuclear Regulatory Commission's (NRC's) Office for Analysis and Evaluation of Operational Data (AEOD) are to: (1) provide feedback of operational experience regarding reactor scram trends in support of the Commission's Strategic Goals, (2) examine the causes of unplanned scrams, and (3) examine the relationship between the causes of unplanned scrams and industry initiatives undertaken to reduce the frequency of unplanned scrams, especially with a view to the potential for future scram rate reduction. 31 refs., 14 figs., 49 tabs

Sub- and super-Maxwellian evaporation of simple gases from liquid water

International Nuclear Information System (INIS)

Kann, Z. R.; Skinner, J. L.

2016-01-01

Non-Maxwellian evaporation of light atoms and molecules (particles) such as He and H 2 from liquids has been observed experimentally. In this work, we use simulations to study systematically the evaporation of Lennard-Jones particles from liquid water. We find instances of sub- and super-Maxwellian evaporation, depending on the mass of the particle and the particle-water interaction strength. The observed trends are in qualitative agreement with experiment. We interpret these trends in terms of the potential of mean force and the effectiveness and frequency of collisions during the evaporation process. The angular distribution of evaporating particles is also analyzed, and it is shown that trends in the energy from velocity components tangential and normal to the liquid surface must be understood separately in order to interpret properly the angular distributions.

Sub- and super-Maxwellian evaporation of simple gases from liquid water

Energy Technology Data Exchange (ETDEWEB)

Kann, Z. R.; Skinner, J. L., E-mail: skinner@chem.wisc.edu [Theoretical Chemistry Institute and Department of Chemistry, University of Wisconsin, Madison, Wisconsin 53706 (United States)

2016-04-21

Non-Maxwellian evaporation of light atoms and molecules (particles) such as He and H{sub 2} from liquids has been observed experimentally. In this work, we use simulations to study systematically the evaporation of Lennard-Jones particles from liquid water. We find instances of sub- and super-Maxwellian evaporation, depending on the mass of the particle and the particle-water interaction strength. The observed trends are in qualitative agreement with experiment. We interpret these trends in terms of the potential of mean force and the effectiveness and frequency of collisions during the evaporation process. The angular distribution of evaporating particles is also analyzed, and it is shown that trends in the energy from velocity components tangential and normal to the liquid surface must be understood separately in order to interpret properly the angular distributions.

Forest evaporation models: Relationships between stand growth and evaporation

CSIR Research Space (South Africa)

Le Maitre, David C

1997-06-01

Full Text Available The relationships between forest stand structure, growth and evaporation were analysed to determine whether forest evaporation can be estimated from stand growth data. This approach permits rapid assessment of the potential impacts of afforestation...

Evaporation of multicomponent chemical spills: When is liquid phase resistance significant?

International Nuclear Information System (INIS)

Berger, D.; Mackay, D.

1993-01-01

When chemicals are spilled on land or water, it is important to be able to estimate evaporation rates accurately. Conventional models used to predict evaporation rates of multicomponent spills assume that the entire resistance to evaporation lies in the vapor phase. Under certain conditions, an additional liquid phase resistance may be introduced, resulting in retarded evaporation rates. Existing models may thus fail to predict spill behavior accurately. A study is described whose objective is to elucidate the significance of the liquid phase resistance. Evaporation experiments were conducted in which a thin layer of synthetic oil (mineral oil enriched with compounds such as pentane, hexane, toluene, octane, and p-xylene) was exposed to prolonged evaporation in a metal tray at controlled wind speeds. Bulk samples of the spill layer were taken at specific time intervals and their composition was determined by gas chromatographic analysis. The results are compared to those from a theoretical model and to gas stripping experiments. The model is based on the evaporative flux equation incorporating Raoult's law; inputs are the air-oil partition coefficient for each component and the composition of the synthetic oil on a volume and mole fraction basis. The study has enabled the formation of vertical concentration profiles to be examined and liquid phase mass transfer coefficients to be estimated. The results imply that liquid-phase resistance effects are likely to be important for the most volatile components. Contaminated areas may thus continue to be hazardous, even though model predictions indicate otherwise. 7 refs., 3 figs., 2 tabs

Fast reactor sodium systems operation experience and 'leak-before-break' criterion

International Nuclear Information System (INIS)

Ivanenko, V.N.; Zybin, V.A.

1996-01-01

In the paper sodium leakage detection systems used at fast reactors are described. Requirements on their main characteristics (sensitivity, response lime) are formulated. Results of tests are presented on studying the parameters of sodium leak detection systems including experiments on the measurement of size distribution of aerosol particles that have passed through sodium systems thermal insulation after leak initiation. Comparison of these data with dispersion of particles formed at free burning is carried out. Experience of real leaks that occurred at fast reactor sodium systems is analyzed. It has been shown that initiation and development of real leaks do not always follow the theoretical scheme. A substantial role of human factor for sodium systems reliability relative to sodium leaks is stressed. (author)

Mapping energetics of atom probe evaporation events through first principles calculations.

Science.gov (United States)

Peralta, Joaquín; Broderick, Scott R; Rajan, Krishna

2013-09-01

The purpose of this work is to use atomistic modeling to determine accurate inputs into the atom probe tomography (APT) reconstruction process. One of these inputs is evaporation field; however, a challenge occurs because single ions and dimers have different evaporation fields. We have calculated the evaporation field of Al and Sc ions and Al-Al and Al-Sc dimers from an L1₂-Al₃Sc surface using ab initio calculations and with a high electric field applied to the surface. The evaporation field is defined as the electric field at which the energy barrier size is calculated as zero, corresponding to the minimum field that atoms from the surface can break their bonds and evaporate from the surface. The evaporation field of the surface atoms are ranked from least to greatest as: Al-Al dimer, Al ion, Sc ion, and Al-Sc dimer. The first principles results were compared with experimental data in the form of an ion evaporation map, which maps multi-ion evaporations. From the ion evaporation map of L1₂-Al₃Sc, we extract relative evaporation fields and identify that an Al-Al dimer has a lower evaporation field than an Al-Sc dimer. Additionally, comparatively an Al-Al surface dimer is more likely to evaporate as a dimer, while an Al-Sc surface dimer is more likely to evaporate as single ions. These conclusions from the experiment agree with the ab initio calculations, validating the use of this approach for modeling APT energetics. Copyright © 2013 Elsevier B.V. All rights reserved.

Evaporative cooling: Effective latent heat of evaporation in relation to evaporation distance from the skin

NARCIS (Netherlands)

Havenith, G.; Bröde, P.; Hartog, E.A. den; Kuklane, K.; Holmer, I.; Rossi, R.M.; Richards, M.; Farnworth, B.; Wang, X.

2013-01-01

Calculation of evaporative heat loss is essential to heat balance calculations. Despite recognition that the value for latent heat of evaporation, used in these calculations, may not always reflect the real cooling benefit to the body, only limited quantitative data on this is available, which has

Vibration behavior of PWR reactor internals Model experiments and analysis

International Nuclear Information System (INIS)

Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

1975-01-01

In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

Overview of the FUTURIX-FTA Irradiation Experiment in the Phénix Reactor

Energy Technology Data Exchange (ETDEWEB)

Heather J.M. Chichester; Steve L. Hayes; Kenneth J. McClellan; Jean-Luc Paul; Marc Masson; Stewart L. Voit; Fabienne Delage

2015-09-01

The Advanced Fuels Campaign utilizes the Advanced Test Reactor (ATR) for most of its irradiation testing. Cadmium-shrouded baskets are used in ATR to modify the neutron spectrum to simulate a fast reactor environment for the fuel. FUTURIX-FTA is an irradiation experiment conducted in the Phenix fast reactor in France. Results from FUTURIX-FTA and irradiation tests in ATR using identical fuel compositions will be compared to identify and evaluate any differences in fuel behavior due to differences in the irradiation source.

French experience in design, operation and revamping of nuclear research reactors, in support of advanced reactors development

International Nuclear Information System (INIS)

Barre, B.; Bergeonneau, P.; Merchie, F.; Minguet, J.L.; Rousselle, P.

1996-01-01

The French nuclear program is strongly based on the R and D work performed in the CEA nuclear research centers and particularly on the various experimental programs carried out in its research reactors in the frame of cooperative actions between the Commissariat a l'Energie Atomique (CEA), Framatome and Electricite de France (EDF). Several types of research reactors have been built by Technicatome and CEA to carry out successfully this considerable R and D work on fuels and materials, among them the socalled Materials Testing Reactors (MTR) SILOE (35 MW) and OSIRIS (70 MW) which are indeed very well suited for technological irradiations. Their simple and flexible design and the large irradiation space available around the core, the SILOE and OSIRIS reactors can be shared by several types of applications such as fuel and material testings for nuclear power plants, radioisotopes production, silicon doping and fundamental research. It is worthwhile recalling that Technicatome and CEA have also built research reactors fully dedicated to safety experimental studies, such as the CABRI, SCARABEE and PHEBUS reactors at Cadarache, and others dedicated to fundamental research such as ORPHEE (14 MW) and the Reacteur a Haut Flux -High Flux Reactor- (RHF 57 MW). This paper will present some of the most significant conceptual and design features of all these reactors as well as the main improvements brought to most of them in the last years. Based on this wide experience, CEA and Technicatome have specially designed for export a new multipurpose research reactor named SIRIUS, with two versions depending on the utilization spectrum and the power range (5 MW to 30 MW). At last, CEA has recently launched the preliminary project study of a new MTR, the Jules Horowitz Reactor, to meet the future needs of fuels and materials irradiations in the next 4 or 5 decades, in support of the French long term nuclear power program. (J.P.N.)

China: EDF's feedback experience of reactor operating is essential to win international markets

International Nuclear Information System (INIS)

Maillart, H.

2016-01-01

The main assets of EDF on the Chinese nuclear power market is first, its very important feedback experience of reactor operations (EDF cumulates one year of reactor operations every week due to its fleet of 58 reactors), secondly the cooperation with China allowed China to enter nuclear energy in 1983 with the construction of the Daya Bay plant and now to develop its own technology: the CPR-1000 reactor. China is the world leader in terms of nuclear market dynamism with 30 reactors in operation, 24 reactors being built and 40 others planned. A new stage in the Franco-China cooperation would be to share relevant good practices in the managing of both French and Chinese fleets of reactors. EDF has upgraded its commercial international offer, it now proposes to cover all the stages of the nuclear power plant from site selection to plant deconstruction via construction, operation, maintenance and waste management which constitutes a commitment over a 100 year period. (A.C.)

Introducing ultrasonic falling film evaporator for moderate temperature evaporation enhancement.

Science.gov (United States)

Dehbani, Maryam; Rahimi, Masoud

2018-04-01

In the present study, Ultrasonic Falling Film (USFF), as a novel technique has been proposed to increase the evaporation rate of moderate temperature liquid film. It is a proper method for some applications which cannot be performed at high temperature, such as foodstuff industry, due to their sensitivity to high temperatures. Evaporation rate of sodium chloride solution from an USFF on an inclined flat plate compared to that for Falling Film without ultrasonic irradiation (FF) at various temperatures was investigated. The results revealed that produced cavitation bubbles have different effects on evaporation rate at different temperatures. At lower temperatures, size fluctuation and collapse of bubbles and in consequence induced physical effects of cavitation bubbles resulted in more turbulency and evaporation rate enhancement. At higher temperatures, the behavior was different. Numerous created bubbles joined together and cover the plate surface, so not only decreased the ultrasound vibrations but also reduced the evaporation rate in comparison with FF. The highest evaporation rate enhancement of 353% was obtained at 40 °C at the lowest Reynolds number of 250. In addition, the results reveal that at temperature of 40 °C, USFF has the highest efficiency compared to FF. Copyright © 2017 Elsevier B.V. All rights reserved.

Experiment monitoring system of a new electromagnet drive for nuclear reactor control rod

International Nuclear Information System (INIS)

Zhang Jige; Wang Xiaoguang; Wu Yuanqiang; Zhang Zhengming

2003-01-01

In order to deal with some unsolved problems in the engineering prototype design of a new electromagnet drive device for nuclear reactor control rod, the property experiment in view of principle prototype is carried out. Actual displacement of nuclear reactor control rod is measured by means of raster ruler and the test data is obtained by means of computer. The computer communicates with PLC using RS232 serial port. The experimental results show that the monitoring system have the properties of high reliability and high precision, and ensures the experiment to accomplish successfully

Increase in the amount of evaporator concentrate from nuclear power plants in cemented products

International Nuclear Information System (INIS)

Costa, Bruna S.; Tello, Clédola C.O.

2017-01-01

Nuclear power plants, research centers and other nuclear facilities are sources of radioactive liquid waste generation. These wastes can come from cooling of the primary reactor system, cleaning spent pool of fuel, washing contaminated clothing, among others. One of the most used methods for the treatment of these aqueous flows is the evaporation, which generates the concentrate of the evaporator, waste classified as low and medium level of radiation. Norms determine that radioactive waste must be minimized, and that to be accepted in repositories, they must be solidified. The work sought to reduce the volume of the evaporated concentrate waste and its subsequent solidification in cement. In order to carry out the tests, the evaporator concentrate (CE) simulation solution was prepared and then dried in an oven. Subsequently, cementation of the dry material was made using cement, fluidizer, NaOH and water. After a curing time of 28 days, the compressive strength tests were made for all specimens obtained, and for the samples that obtained resistance above that required by the norm, which is 10MPa, the percentages of reject incorporated and volume reduction. The results showed that, by drying the evaporator concentrate, it was possible to reduce the volume of the waste generated by up to 27% in relation to the waste without drying, which shows that drying is an effective way to increase the incorporation of the evaporator concentrate in packaged waste

. Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

International Nuclear Information System (INIS)

Yusuf, I; Mati, A. A.

2010-01-01

The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

Mass hierarchy sensitivity of medium baseline reactor neutrino experiments with multiple detectors

Energy Technology Data Exchange (ETDEWEB)

Wang, Hong-Xin, E-mail: hxwang@iphy.me [Department of Physics, Nanjing University, Nanjing 210093 (China); Zhan, Liang; Li, Yu-Feng; Cao, Guo-Fu [Institute of High Energy Physics, Chinese Academy of Sciences, Beijing 100049 (China); Chen, Shen-Jian [Department of Physics, Nanjing University, Nanjing 210093 (China)

2017-05-15

We report the neutrino mass hierarchy (MH) determination of medium baseline reactor neutrino experiments with multiple detectors, where the sensitivity of measuring the MH can be significantly improved by adding a near detector. Then the impact of the baseline and target mass of the near detector on the combined MH sensitivity has been studied thoroughly. The optimal selections of the baseline and target mass of the near detector are ∼12.5 km and ∼4 kton respectively for a far detector with the target mass of 20 kton and the baseline of 52.5 km. As typical examples of future medium baseline reactor neutrino experiments, the optimal location and target mass of the near detector are selected for the specific configurations of JUNO and RENO-50. Finally, we discuss distinct effects of the reactor antineutrino energy spectrum uncertainty for setups of a single detector and double detectors, which indicate that the spectrum uncertainty can be well constrained in the presence of the near detector.

Implementation of a model reference adaptive control system using neural network to control a fast breeder reactor evaporator

International Nuclear Information System (INIS)

Ugolini, D.; Yoshikawa, S.; Endou, A.

1994-01-01

Artificial intelligence is foreseen as the base for new control systems aimed to replace traditional controllers and to assist and eventually advise plant operators. This paper discusses the development of an indirect model reference adaptive control (MRAC) system, using the artificial neural network (ANN) technique, and its implementation to control the outlet steam temperature of a sodium to water evaporator. The ANN technique is applied in the identification and in the control process of the indirect MRAC system. The emphasis is placed on demonstrating the efficacy of the indirect MRAC system in controlling the outlet steam temperature of the evaporator, and on showing the important function covered by the ANN technique. An important characteristic of this control system is that it relays only on some selected input variables and output variables of the evaporator model. These are the variables that can be actually measured or calculated in a real environment. The results obtained applying the indirect MRAC system to control the evaporator model are quite remarkable. The outlet temperature of the steam is almost perfectly kept close to its desired set point, when the evaporator is forced to depart from steady state conditions, either due to the variation of some input variables or due to the alteration of some of its internal parameters. The results also show the importance of the role played by the ANN technique in the overall control action. The connecting weights of the ANN nodes self adjust to follow the modifications which may occur in the characteristic of the evaporator model during a transient. The efficiency and the accuracy of the control action highly depends on the on-line identification process of the ANN, which is responsible for upgrading the connecting weights of the ANN nodes. (J.P.N.)

Development of the “Approach to Critical” Experiment Simulation Model for the Consort Reactor Using LABVIEW

International Nuclear Information System (INIS)

Abbo, D. N. O.

2015-01-01

Following the shutdown of the CONSORT reactor, the “Approach to Critical” experiment which allowed students to observe and understand the procedure for taking the reactor to critical, balancing the system at low power and increasing the power over a range of powers levels and eventual reactor shutdown, would no longer be possible. It was therefore important to develop a simulation model of the experiment that would enable future students to have comparable training. An “Approach to Critical” Experiment Simulation model for the CONSORT Reactor was developed using Lab-VIEW software to simulate the “Students” experiment version. Lab– VIEW software was chosen due to its good user graphical user interface, offers ready to start functions and also the possibility of improving on the system with new algorithms. The modulation process was used to develop mathematical codes from equations using Lab–VIEW 2012 based on the CONSORT historical experimental data and known literature. The Simulation models the kinetics of a sub-critical reactor with a start-up neutron source, such that control rods are used to increase the power, then achieve power balance and finally shutting down the reactor. Reactivity changes due to temperature effects were neglected. The model was validated by testing the code through performing the three parts of the experiment; Approach to Critical, Doubling time method and Rod drop method, and results compared to the historical experimental data. The results were in agreement with historical data. However the negligible variations were obtained in the Rod drop method due to the reactivity values used to generate the code. (author)

The Safe and Efficient Evaporation of a Solvent from Solution

Science.gov (United States)

Mahon, Andrew R.

1997-02-01

The process of evaporating a solvent from a solution can cause problems for many students. By using a water-vacuum aspirator, backflashes of water can flood the sample tube and be detrimental to the experiment. This type of apparatus can also cause problems by drawing the solution it is evaporating back into the vacuum hose, causing the student to lose part or all of the products of their experiment. Macroscale and Microscale Organic Experiments, 2nd edition (1), suggested two techniques to dissolve solvents from a mixture. It suggested blowing a stream of air over the solution from a Pasteur pipet, or attaching a Pasteur pipet to an aspirator and drawing air over the surface of the liquid. Again, the danger of blowing air over the solution leaves the risk of splattering the solution, and drawing air over the surface of the liquid as described further endangers the products of the experiment through the risk of sucking the products up into the pipet aspirator. In an effort to eliminate these problems, a new technique has been developed. By inverting an ordinary 200-mL vacuum flask and pulling a steady current of air from the vacuum apparatus through it, any type of small container can be placed under it, allowing the solvent to be evaporated in a steady, mistake-free manner . By evaporating the solvent from the container that the products will be submitted in, no sample is lost through the process of transferring it from a vacuum flask or beaker to the final container.

Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

International Nuclear Information System (INIS)

Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

1986-05-01

In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant

Basic experiments of reactor physics using the critical assembly TCA

International Nuclear Information System (INIS)

Obara, Toru; Igashira, Masayuki; Sekimoto, Hiroshi; Nakajima, Ken; Suzaki, Takenori.

1994-02-01

This report is based on lectures given to graduate students of Tokyo Institute of Technology. It covers educational experiments conducted with the Tank-Type Critical Assembly (TCA) at Japan Atomic Energy Research Institute in July, 1993. During this period, the following basic experiments on reactor physics were performed: (1) Critical approach experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, (5) Measurement of safety sheet worth by the rod drop method. The principle of experiments, experimental procedure, and analysis of results are described in this report. (author)

Experience with reactor assembly of FBTR

International Nuclear Information System (INIS)

Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

2006-01-01

Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

Influence of surface wettability on transport mechanisms governing water droplet evaporation.

Science.gov (United States)

Pan, Zhenhai; Weibel, Justin A; Garimella, Suresh V

2014-08-19

Prediction and manipulation of the evaporation of small droplets is a fundamental problem with importance in a variety of microfluidic, microfabrication, and biomedical applications. A vapor-diffusion-based model has been widely employed to predict the interfacial evaporation rate; however, its scope of applicability is limited due to incorporation of a number of simplifying assumptions of the physical behavior. Two key transport mechanisms besides vapor diffusion-evaporative cooling and natural convection in the surrounding gas-are investigated here as a function of the substrate wettability using an augmented droplet evaporation model. Three regimes are distinguished by the instantaneous contact angle (CA). In Regime I (CA ≲ 60°), the flat droplet shape results in a small thermal resistance between the liquid-vapor interface and substrate, which mitigates the effect of evaporative cooling; upward gas-phase natural convection enhances evaporation. In Regime II (60 ≲ CA ≲ 90°), evaporative cooling at the interface suppresses evaporation with increasing contact angle and counterbalances the gas-phase convection enhancement. Because effects of the evaporative cooling and gas-phase convection mechanisms largely neutralize each other, the vapor-diffusion-based model can predict the overall evaporation rates in this regime. In Regime III (CA ≳ 90°), evaporative cooling suppresses the evaporation rate significantly and reverses entirely the direction of natural convection induced by vapor concentration gradients in the gas phase. Delineation of these counteracting mechanisms reconciles previous debate (founded on single-surface experiments or models that consider only a subset of the governing transport mechanisms) regarding the applicability of the classic vapor-diffusion model. The vapor diffusion-based model cannot predict the local evaporation flux along the interface for high contact angle (CA ≥ 90°) when evaporative cooling is strong and the

Falling film evaporators: organic solvent regeneration in nuclear fuel reprocessing

International Nuclear Information System (INIS)

Garcin, I.

1989-01-01

The aim of this work was to improve knowledge about working of falling film evaporators used in nuclear fuel reprocessing plants for organic solvent regeneration. The first part deals with a non evaporation film. An original film thickness measuring technique was used; infrared thermography. It gave indications on hydrodynamics and wave amplitude and pointed out thermocapillary forces to be the cause of bad wetting of the heated wall. By another way we showed that a small slit spacing on the film distributor, an enhanced surface roughness and an important liquid flow rate favour a better wetting. The second part deals with evaporation of a binary solvent mixture. Experiments in an industrial evaporator corroborated the fact that it is essential for the efficiency of the apparatus to work at high flow rates. We propose an over-simple model which can be used to estimate performances of co-current falling film evaporators of the process [fr

Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

International Nuclear Information System (INIS)

Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

2013-01-01

The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

Determination of the neutron flux for the Yankee Rowe experiment in the Ford Nuclear Reactor

International Nuclear Information System (INIS)

Cacciapouti, R.J.; Petrusha, L.

1994-01-01

Yankee Atomic Electric Company undertook a Test Irradiation Program at the Ford Nuclear Reactor of the University of Michigan. The program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials. The program was also intended to remove uncertainties in the existing reactor vessel fluence and damage predictions on the Yankee Rowe reactor vessel steel. Since this is the first in-core experiment of this type for the Ford Nuclear Reactor, the measurement of the reaction rate and the estimate of the fluence are presented

Artificial weathering of oils by rotary evaporator

International Nuclear Information System (INIS)

Fieldhouse, B.; Hollebone, B.P.; Singh, N.R.; Tong, T.S.; Mullin, J.

2009-01-01

Oil weathering has a considerable affect on the behaviour, impact and ultimate fate of an oil spill. As such, efforts have been made to study weathering as a whole using bench-scale procedures. The studies are generally divided into individual processes where the effect of other major processes are introduce as an amended sample input rather than a concurrent process. The weathering process that has the greatest effect immediately following an oil spill is evaporation, particularly for lighter oils. The rotary evaporator apparatus offers a convenient means of producing artificially weathered oil for laboratory studies. This paper reported on a study that examined the representativeness of samples obtained by this method compared to pan evaporation and the impact of changes to the apparatus or method parameters on sample chemistry. Experiments were performed on Alberta Sweet Mixed Blend no. 5 in a rotary evaporator under varying conditions of temperature and air flow at ambient pressure using 2 apparatus. The rate of mass loss increased with temperature and air flow rate as expected, but the quantitative relationships could not be defined from the data due to contributions by other uncontrolled factors. It was concluded that the rotary evaporator is not suited for evaporation rate studies, but rather for producing samples suitable for use in other studies. Chemical analysis showed that the relative abundance distributions of target n-alkane hydrocarbons varied with the degree of weathering of an oil in a consistent manner at ambient pressure, regardless of the temperature, rate of air exchange or other factors related to the apparatus and procedure. The composition of the artificially weathered oil was also consistent with that from an open pan simulation of a weathered oil slick. Loss of water content varied with the conditions of evaporation because of the differential rates of evaporation due to relative humidity considerations. It was concluded that weathering

Vacuum evaporation of pure metals

OpenAIRE

Safarian, Jafar; Engh, Thorvald Abel

2013-01-01

Theories on the evaporation of pure substances are reviewed and applied to study vacuum evaporation of pure metals. It is shown that there is good agreement between different theories for weak evaporation, whereas there are differences under intensive evaporation conditions. For weak evaporation, the evaporation coefficient in Hertz-Knudsen equation is 1.66. Vapor velocity as a function of the pressure is calculated applying several theories. If a condensing surface is less than one collision...

Evaporation in hydrology and meteorology

OpenAIRE

Brandsma, T.

1990-01-01

In this paper the role of evaporation in hydrology and meteorology is discussed, with the emphasis on hydrology. The basic theory of evaporation is given and methods to determine evaporation are presented. Some applications of evaporation studies in literature are given in order to illustrate the theory. Further, special conditions in evaporation are considered, followed by a fotmulation of the difficulties in determining evaporation, The last part of the paper gives a short discussion about ...

The RERTR demonstration experiments program at the Ford Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Wehe, D K; King, J S [Department of Nuclear Engineering, University of Michigan (United States)

1983-08-01

The purpose of this paper is to highlight a major part of the experimental work which is being carried out at the Ford Nuclear Reactor (FNR) in conjunction with the RERTR program. A demonstration experiments program has been developed to: 1) characterize the FNR in sufficient detail to discern and quantify neutronic differences between the high and low enriched cores; 2) provide the theoretical group with measurements to benchmark their calculations. As with any experimental program associated with a reactor, stringent constraints limit the experiments which can be performed. Some experiments are performed routinely on the FNR (such as control rod calibrations), and much data is already available. Unfortunately, the accuracy we demand precludes using much of this earlier data. And in many cases, the requirement of precise (and copious) data has led to either developing new techniques (as in the case of rhodium mapping and neutron diffraction) or to further refinements on existing methods (as in the case of spectral unfolding). Nevertheless, we have tried to stay within the realm of recognized, well-established experimental methods in order to assuage any doubts about measured differences between HEU and LEU core parameters. This paper describes the principal results of the experiments performed so far.

Micromodel observations of evaporative drying and salt deposition in porous media

Science.gov (United States)

Rufai, Ayorinde; Crawshaw, John

2017-12-01

Most evaporation experiments using artificial porous media have focused on single capillaries or sand packs. We have carried out, for the first time, evaporation studies on a 2.5D micromodel based on a thin section of a sucrosic dolomite rock. This allowed direct visual observation of pore-scale processes in a network of pores. NaCl solutions from 0 wt. % (de-ionized water) to 36 wt. % (saturated brine) were evaporated by passing dry air through a channel in front of the micromodel matrix. For de-ionized water, we observed the three classical periods of evaporation: the constant rate period (CRP) in which liquid remains connected to the matrix surface, the falling rate period, and the receding front period, in which the capillary connection is broken and water transport becomes dominated by vapour diffusion. However, when brine was dried in the micromodel, we observed that the length of the CRP decreased with increasing brine concentration and became almost non-existent for the saturated brine. In the experiments with brine, the mass lost by evaporation became linear with the square root of time after the short CRP. However, this is unlikely to be due to capillary disconnection from the surface of the matrix, as salt crystals continued to be deposited in the channel above the matrix. We propose that this is due to salt deposition at the matrix surface progressively impeding hydraulic connectivity to the evaporating surface.

Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

International Nuclear Information System (INIS)

Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

1982-01-01

Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

NEA activities in preserving, evaluating and applying data from fast reactor experiments

International Nuclear Information System (INIS)

Gulliford, Jim; Cornet, S.M.; Hill, I.; Yamaji, A.

2013-01-01

Conclusions: Progress to date: • Extensive programme of work to preserve and evaluate data from integral experiments has been established since the mid 1990s. • NEA Data Bank maintains and distributes several databases of these integral experiments, notably through the ICSBEP and IRPhE projects. • More recently programmes of work have been established to help preserve data from the UK Fast Reactor Programme and from various experiments related to minor actinide management. • Data obtained from these programmes are made available to the nuclear science community to provide high quality benchmarks against which modelling methods can be validated. • Involvement of younger scientists and engineers to work alongside well-established experts in the process of evaluating the information is a highly efficient means of transmitting tacit knowledge to the new generation of nuclear specialists. Conclusions: Looking ahead - • Further development of Databases and Database tools, e.g. – improved coverage of fast reactor experiments, MAs; – improved treatment of correlations in uncertainties between experiments; – production of sensitivities to facilitate identification of similar experiments. • Continuation securing UK archives and creating framework for information: – Start identifying suitable integral experiments for inclusion in NEA databases

Liquid radioactive waste concentration by the method of evaporation from porous plates

International Nuclear Information System (INIS)

Dmitriev, S.A.; Karlin, Yu.V.; Maryakhin, M.A.; Myasnikov, Yu.G.; Slastennikov, Yu.T.

2009-01-01

As it is shown by bench-scale experiments radioactive effluents are concentrated to salt content 319 g/l at temperature lower, than evaporation temperature of water, and specific power inputs lower, than specific evaporation heat of water by 20 times. Results of tests at pilot plant (productivity to 43 kg/h by evaporation water) that is placed in mobile water purification unit ECO are described. This unit is used for radioactive water treatment from different organizations at SPU Radon

Proceedings of US-Japan workshop on new generation experiments and reactors (joined by EC)

International Nuclear Information System (INIS)

1988-07-01

The workshop, titled 'New Generation Experiments and Reactors', was held at Plasma Physics Laboratory, Kyoto University from 25 to 28 July 1988. The purpose of the meeting was to review the latest achievements and status of stellarator/heliotron new generation experiments as well as the prospects for stellarator/heliotron fusion reactors on the occasion when the New Large Helical System of MOE in Japan is being realized. The reports on the New Large Helical System of MOE cover an overview, physics issues, design, MHD studies, transport code results and bootstrap current, particle orbit studies, divertor studies, NBI heating, analysis of wave heating, heating system, diagnostics, and SC coil technology. The reports on ATF II cover an overview, physics studies strategy, status of physics studies, engineering issues, perspective of helical systems, issues for next-generation experiments and relationship to the Univ. of Wisconsin Program, and issues for next-generation experiments and relation to Auburn Program. Other reports address recent studies of present devices, studies related with WVIIX, and reactor studies. (N.K.)

UK experience of safety requirements for thermal reactor stations

International Nuclear Information System (INIS)

Matthews, R.R.; Dale, G.C.; Tweedy, J.N.

1977-01-01

The paper summarises the development of safety requirements since the first of the Generating Boards' Magnox reactors commenced operation in 1962 and includes A.G.R. safety together with the preparation of S.G.H.W.R. design safety criteria. It outlines the basic principles originally adopted and shows how safety assessment is a continuing process throughout the life of a reactor. Some description is given of the continuous effort over the years to obtain increased safety margins for existing and new reactors, taking into account the construction and operating experience, experimental information, and more sophisticated computer-aided design techniques which have become available. The main safeguards against risks arising from the Generating Boards' reactors are the achievement of high standards of design, construction and operation, in conjunction with comprehensive fault analyses to ensure that adequate protective equipment is provided. The most important analyses refer to faults which can lead to excessive fuel element temperatures arising from an increase in power or a reduction in cooling capacity. They include the possibility of unintended control rod withdrawal at power or at start-up, coolant flow failure, pressure circuit failure, loss of boiler feed water, and failure of electric power. The paper reviews the protective equipment, and the policy for reactor safety assessments which include application of maximum credible accident philosophy and later the limited use of reliability and probability methods. Some of the Generating Boards' reactors are now more than half way through their planned working lives and during this time safety protective equipment has occasionally been brought into operation, often for spurious reasons. The general performance, of safety equipment is reviewed particularly for incidents such as main turbo-alternator trip, circulator failure, fuel element failures and other similar events, and some problems which have given rise to

Some scoping experiments for a space reactor

International Nuclear Information System (INIS)

Alexander, C.A.; Ogden, J.S.

1983-01-01

Some scoping experiments were performed to evaluate fuel performance in a lithium heat pipe reactor operating at a nominal 1500K heat pipe temperature. Fuel-coolant and fuel-coolant-clad relationships showed that once a failed heat pipe occurs temperatures can rise high enough so that large concentrations of uranium can be transported by the vapor phase. Upon condensation this uranium would be capable of penetrating heat pipes adjacent to the failed pipe. The potential for propagation of failure exists with UO 2 and a lithium heat pipe. Changing the composition of the metal of the heat pipe would have only a second order effect on the kinetics of the failure mechanism. Uranium carbide and nitride were considered as potential fuels which are nonreactive in a lithium environment. At high temperatures the nitride would be favored because of its better compatibility with potential cladding materials. Compositions of UN with small additions of YN appear to offer very attractive properties for a compact high temperature high power density reactor

On the link between potential evaporation and regional evaporation from a CBL perspective

Science.gov (United States)

Lhomme, J. P.; Guilioni, L.

2010-07-01

The relationship between potential evaporation and actual evaporation was first examined by Bouchet (Proc Berkeley Calif Symp IAHS Publ, 62:134-142, 1963) who considered potential evaporation as the consequence of regional evaporation due to atmospheric feedbacks. Using a heuristic approach, he derived a complementary relationship which, despite no real theoretical background, has proven to be very useful in interpreting many experimental data under various climatic conditions. Here, the relationship between actual and potential evaporation is reinterpreted in the context of the development of the convective boundary layer (CBL): first, with a closed-box approach, where the CBL has an impermeable lid; and then with an open system, where air is exchanged between the CBL and its external environment. By applying steady forcing to these systems, it is shown that an equilibrium state is reached, where potential evaporation has a specific equilibrium formulation as a function of two parameters: one representing large-scale advection and the other the feedback effect of regional evaporation on potential evaporation, i.e. a kind of “medium-scale advection”. It is also shown that the original form of Bouchet’s complementary relationship is not verified in the equilibrium state. This analysis leads us to propose a new and more rational approach of the relationship between potential and actual evaporation through the effective surface resistance of the region.

Trends and experiences in reactor coolant pump motors

International Nuclear Information System (INIS)

Anon.

1980-01-01

A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

The results of reactor material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address exvessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debrids characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity. (orig.)

Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

International Nuclear Information System (INIS)

Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

1984-01-01

Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

Evaporation of liquids on chemically patterned surfaces

NARCIS (Netherlands)

Vieyra Salas, J.A.; Darhuber, A.A.

2011-01-01

We studied evaporation rates of volatile liquids deposited onto chemically patterned surfaces by means of experiments and numerical simulations. We quantified the influence of the droplet geometry, in particular circular, triangular, rectangular and square shapes, as well as the influence of contact

EVAPORATION FORM OF ICE CRYSTALS IN SUBSATURATED AIR AND THEIR EVAPORATION MECHANISM

OpenAIRE

ゴンダ, タケヒコ; セイ, タダノリ; Takehiko, GONDA; Tadanori, SEI

1987-01-01

The evaporation form and the evaporation mechanism of dendritic ice crystals grown in air of 1.0×(10)^5 Pa and at water saturation and polyhedral ice crystals grown in air of 4.0×10 Pa and at relatively low supersaturation are studied. In the case of dendritic ice crystals, the evaporation preferentially occurs in the convex parts of the crystal surfaces and in minute secondary branches. On the other hand, in the case of polyhedral ice crystals, the evaporation preferentially occurs in the pa...

Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

International Nuclear Information System (INIS)

Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

1982-01-01

During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

The radon monitoring system in Daya Bay Reactor Neutrino Experiment

International Nuclear Information System (INIS)

Chu, M.C.; Kwan, K.K.; Kwok, M.W.; Kwok, T.; Leung, J.K.C.; Leung, K.Y.; Lin, Y.C.; Luk, K.B.; Pun, C.S.J.

2016-01-01

We developed a highly sensitive, reliable and portable automatic system (H 3 ) to monitor the radon concentration of the underground experimental halls of the Daya Bay Reactor Neutrino Experiment. H 3 is able to measure radon concentration with a statistical error less than 10% in a 1-h measurement of dehumidified air (R.H. 5% at 25 °C) with radon concentration as low as 50 Bq/m 3 . This is achieved by using a large radon progeny collection chamber, semiconductor α-particle detector with high energy resolution, improved electronics and software. The integrated radon monitoring system is highly customizable to operate in different run modes at scheduled times and can be controlled remotely to sample radon in ambient air or in water from the water pools where the antineutrino detectors are being housed. The radon monitoring system has been running in the three experimental halls of the Daya Bay Reactor Neutrino Experiment since November 2013.

Field experience in use of radiation instruments in Cirus reactor

International Nuclear Information System (INIS)

Ramesh, N.; Sharma, R.C.; Agarwal, S.K.; Sawant, D.K.; Yadav, R.K.B.; Prasad, S.K.

2005-01-01

Cirus, located at Bhabha Atomic Research Centre, is a 40 MW (Th) research reactor fuelled by natural uranium, moderated by heavy water and cooled by de-mineralized light water. Graphite is used as reflector in this reactor. The reactor, commissioned in the year 1960, was in operation with availability factor of about 70% till early nineties. There after signs of ageing started surfacing up. After ageing studies, refurbishment plan was finalized and executed during the period from 1997-2002. after successful refurbishment, the reactor is in operation at full power. A wide range of radiation instruments have been used at Cirus for online monitoring of the radiological status of various process systems and environmental releases. Also, variety of survey meters, counting systems and monitors have been used by the health physics unit of the reactor for radiation hazard control. Many of these instruments, which were originally of Canadian design, have undergone changes due to obsolescence or as part of upgradation. This paper describes the experience with the radiation instruments of Cirus, bringing out their effectiveness in meeting the design intent, difficulties faced in their field use, and modifications carried out based on the performance feed back. Also, this paper highlights the areas where further efforts are needed to develop nuclear instrumentation to further strengthen monitoring and surveillance. (author)

Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

International Nuclear Information System (INIS)

Tomberlin, T.A.

2002-01-01

The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed

Prediction of heat and mass transfer in innovative nuclear reactors

International Nuclear Information System (INIS)

Ambrosini, W.; Forgione, N.; Manfredini, A.; Oriolo, F.

2000-01-01

This paper proposes a short review of the different forms adopted to express the analogy between heat and mass transfer for application in correlating data from condensation and evaporation experiments. In particular, the assumptions at the basis of the various forms presented by classical textbooks as well as recent research work are qualitatively discussed, proposing a unified treatment of the different models. On this background, the results of the application of one of the considered forms of the analogy to a problem having relevance for nuclear reactor safety are then discussed. The work performed in this frame is related to condensation on finned tube heat exchangers, proposed as key components in passive containment cooling systems adopted in some innovative reactor concepts. The application of the model to the experimental dana also allowed to obtain interesting information about the effect of different parameters on the cooling capabilities of this compact heat exchangers. (author)

Blast from explosive evaporation of carbon dioxide : Experiment, modeling and physics

NARCIS (Netherlands)

Van der Voort, M.M.; Van den berg, A.C.; Roekaerts, D.J.E.M.; Xie, M.; De Bruijn, P.C.J.

2012-01-01

Explosive evaporation of a superheated liquid is a relevant hazard in the process industry. A vessel rupture during storage, transport or handling may lead to devastating blast effects. In order to assess the risk associated with this hazard or to design protective measures, an accurate prediction

Blast from explosive evaporation of carbon dioxide: Experiment, modeling and physics

NARCIS (Netherlands)

Voort, M.M. van der; Berg, A.C. van den; Roekaerts, D.J.E.M.; Xie, M.; Bruijn, P.C.J. de

2012-01-01

Explosive evaporation of a superheated liquid is a relevant hazard in the process industry. A vessel rupture during storage, transport or handling may lead to devastating blast effects. In order to assess the risk associated with this hazard or to design protective measures, an accurate prediction

Small scale thermal-hydraulic experiment for stable operation of a pius-type reactor

International Nuclear Information System (INIS)

Tasaka, K.; Tamaki, M.; Imai, S.; Irianto, I.D.; Tsuji, Y.; Kukita, Y.

1994-01-01

Thermal-hydraulic experiments using a small-scale atmospheric pressure test loop have been performed for the Process Inherent Ultimate Safety (PIUS)-type reactor to develop the new pump speed feedback control system. Three feedback control systems based on the measurement of flow rate, differential pressure, and fluid temperature distribution in the lower density lock have been proposed and confirmed by a series of experiments. Each of the feedback control systems had been verified in the simulation experiment such as a start-up simulation test. The automatic pump speed control based on the fluid temperature at the lower density lock was quite effective to maintain the stratified interface between primary water and borated pool water for stable operation of the reactor. (author)

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowell, B.S.; Fisher, S.E.

1999-02-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

International Nuclear Information System (INIS)

Cowell, B.S.; Fisher, S.E.

1999-01-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

'Experience with decommissioning of research and test reactors at Argonne National Laboratory'

International Nuclear Information System (INIS)

Bhattacharyya, S.K.; Yule, T.J.; Fellhauer, C.R.; Boing, L.E.

2002-01-01

A large number of research reactors around the world have reached the end of their useful operational life. Many of these are kept in a controlled storage mode awaiting decontamination and decommissioning (D and D). At Argonne National Laboratory located near Chicago in the United States of America, significant experience has been gained in the D and D of research and test reactors. These experiences span the entire range of activities in D and D - from planning and characterization of the facilities to the eventual disposition of all waste. A multifaceted D nd D program has been in progress at the Argonne National Laboratory - East site for nearly a decade. The program consists of three elements: - D and D of nuclear facilities on the site that have reached the end of their useful life; - Development and demonstrations of technologies that help in safe and cost effective D and D; - Presentation of training courses in D and D practices. Nuclear reactor facilities have been constructed and operated at the ANL-E site since the earliest days of nuclear power. As a result, a number of these early reactors reached end-of-life long before reactors on other sites and were ready for D and D earlier. They presented an excellent set of test beds on which D and D practices and technologies could be demonstrated in environments that were similar to commercial reactors, but considerably less hazardous. As shown, four reactor facilities, plutonium contaminated glove boxes and hot cells, a cyclotron facility and assorted other nuclear related facilities have been decommissioned in this program. The overall cost of the program has been modest relative to the cost of comparable projects undertaken both in the U.S. and abroad. The safety record throughout the program was excellent. Complementing the actual operations, a set of D and D technologies are being developed. These include robotic methods of tool handling and operation, chemical and laser decontamination techniques, sensors

Operating experience and maintenance at the TRIGA Mark II LENA reactor

International Nuclear Information System (INIS)

Cingoli, F.; Altieri, S.; Lana, F.; Rosti, G.; Alloni, L.; Meloni, S.

1988-01-01

The last two years at the Trigs Mark II LENA plant were characterized by the running of the n-n-bar oscillation NADIR experiment. Consequently reactor operation was positively affected and the running hours rose again above 1000 hours per year. The LENA team was also deeply involved in the procedures for the renewal of the reactor operation license. The new requirements set by the Nuclear Energy Licensing Authority (ENEA for Italy) most of which concerning radiation protection and environmental impact, have been already fulfilled. In some cases the installation of new apparatus is underway

Development, operational experience and implications for future design of fast reactors in Western Europe

International Nuclear Information System (INIS)

Brandstetter, A.; Broomfield, A.M.; Saitcevsky, B.

1990-01-01

Over the past 30 years, the partners now collaborating in Europe on fast-reactor development have taken the technology from a theoretical possibility to an engineering reality. In that time there has been a progression from experimental zero-energy facilities followed by small power-producing engineering test reactors, to prototype reactors and a large commercial-size reactor. The paper describes the highlights of the reactor programmes in the partner countries and by example illustrates the experience gained from reactor operation and some of the principal activities in the supporting development programme. These include such topics as fuel performance, fast-neutron physics, liquid-metal thermal hydraulics, sodium chemistry, instrumentation and safety aspects. The paper concludes by summarizing some of the main objectives of the current development programme, which is directed to the support of the European Fast Reactor design being prepared by the European design and construction companies. (author)

Fast reactor operating experience gained in Russia: Analysis of anomalies and abnormal operation cases

International Nuclear Information System (INIS)

Ashurko, Y.M.; Baklushin, R.P.; Zagorulko, Y.I.; Ivanenko, V.N.; Matveyev, V.P.; Vasilyev, B.A.

2000-01-01

Review of various anomalous events and abnormal operation experience gained in the process of Russian fast reactors operation is given in the paper. The main information refers to the BN-600 demonstration reactor operation. Statistical data on sodium leaks and steam generator failures are presented, and sources of these events and countermeasures taken to avoid their appearance on the operating reactors as well as related changes made in the BN-800 reactor design are considered. In the paper, some features of impurities behaviour are considered in various modes of the BN-600 reactor operation. Information is given on the impurities ingress into the circuits, on abnormal situation emerged in the process of the BN-600 reactor operation and its probable cause. Information is presented on the event related to the increased torque of the BN-600 reactor central rotating column and repair works performed. (author)

Model of an Evaporating Drop Experiment

Science.gov (United States)

Rodriguez, Nicolas

2017-11-01

A computational model of an experimental procedure to measure vapor distributions surrounding sessile drops is developed to evaluate the uncertainty in the experimental results. Methanol, which is expected to have predominantly diffusive vapor transport, is chosen as a validation test for our model. The experimental process first uses a Fourier transform infrared spectrometer to measure the absorbance along lines passing through the vapor cloud. Since the measurement contains some errors, our model allows adding random noises to the computational integrated absorbance to mimic this. Then the resulting data are interpolated before passing through a computed tomography routine to generate the vapor distribution. Next, the gradients of the vapor distribution are computed along a given control volume surrounding the drop so that the diffusive flux can be evaluated as the net rate of diffusion out of the control volume. Our model of methanol evaporation shows that the accumulated errors of the whole experimental procedure affect the diffusive fluxes at different control volumes and are sensitive to how the noisy data of integrated absorbance are interpolated. This indicates the importance of investigating a variety of data fitting methods to choose which is best to present the data. Trinity University Mach Fellowship.

Evaporation and Climate Change

NARCIS (Netherlands)

Brandsma, T.

1993-01-01

In this article the influence of climate change on evaporation is discussed. The emphasis is on open water evaporation. Three methods for calculating evaporation are compared considering only changes in temperature and factors directly dependent on temperature. The Penman-method is used to

Educational reactor-physics experiments with the critical assemble TCA

Energy Technology Data Exchange (ETDEWEB)

Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki [Tokyo Inst. of Tech. (Japan); Horiki, Oichiro; Suzaki, Takenori

1997-10-01

The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for (1) Critical approach and Exponential experiment, (2) Measurement of neutron flux distribution, (3) Measurement of power distribution, (4) Measurement of fuel rod worth distribution, and (5) Measurement of safety plate worth by the rod drop method. (author)

Educational reactor-physics experiments with the critical assembly TCA

International Nuclear Information System (INIS)

Tsutsui, Hiroaki; Okubo, Masaaki; Igashira, Masayuki; Horiki, Oichiro; Suzaki, Takenori.

1997-10-01

The Tank-Type Critical Assembly (TCA) of Japan Atomic Energy Research Institute is research equipment for light water reactor physics. In the present report, the lectures given to the graduate students of Tokyo Institute of Technology who participated in the educational experiment course held on 26-30 August at TCA are rearranged to provide useful information for those who will implement educational basic experiments with TCA in the future. This report describes the principles, procedures, and data analyses for 1) Critical approach and Exponential experiment, 2) Measurement of neutron flux distribution, 3) Measurement of power distribution, 4) Measurement of fuel rod worth distribution, and 5) Measurement of safety plate worth by the rod drop method. (author)

Convection-enhanced water evaporation

OpenAIRE

B. M. Weon; J. H. Je; C. Poulard

2011-01-01

Water vapor is lighter than air; this can enhance water evaporation by triggering vapor convection but there is little evidence. We directly visualize evaporation of nanoliter (2 to 700 nL) water droplets resting on silicon wafer in calm air using a high-resolution dual X-ray imaging method. Temporal evolutions of contact radius and contact angle reveal that evaporation rate linearly changes with surface area, indicating convective (instead of diffusive) evaporation in nanoliter water droplet...

HEU to LEU conversion experience at the UMass-Lowell research reactor

International Nuclear Information System (INIS)

White, John R.; Bobek, Leo M.

2005-01-01

The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (author)

Operational Experience with the TRIGA Mark II Reactor of the University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Tigliole, A. Borio Di; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Provasi, M.C.; Salvini, A.; Scian, G.; Vinciguerra, G. [University of Pavia, Laboratory of Applied Nuclear Energy (L.E.N.A), Via Aselli 41, 27100 Pavia (Italy)

2011-07-01

The Laboratory of Applied Nuclear Energy (LENA) is an Interdepartmental Research Centre of the University of Pavia which operates a 250 kW TRIGA Mark II Research Nuclear Reactor, a Cyclotron for the production of radioisotopes and other irradiation facilities. The reactor is in operation since 1965 and many home-made upgrading were realized in the past years in order to assure a continuous operation of the reactor for the future. The annual reactor operational time at nominal power is in the range of 300 - 400 hours depending upon the time schedule of some experiments and research activities. The reactor is mainly used for NAA activities, BNCT research, samples irradiation and training. In specific, few tens of hours of reactor operation per year are dedicated to training courses for University students and for professionals. Besides, the LENA Centre hosts every year more than one thousand high school students in visit. Lately, LENA was certified ISO 9001:2008 for the ''operation and maintenance of the reactor'' and for the ''design and delivery of the irradiation service''. Nowadays the reactor shows a good technical state and, at the moment, there are no political or economical reason to consider the reactor shut-down. (author)

Benefits of reactor physics experiments for the HTGR industrial development - an attempt to a quantitative approach

Energy Technology Data Exchange (ETDEWEB)

Cuniberti, R; Graziani, G; Massino, L; Rinaldini, C; Zanantoni, C

1972-10-15

The available results of reactor physics experiments on HTGRs and their accuracies are briefiy reviewed. The physical quantities of interest are grouped into three categories: basic nuclear data, lattice parameters and integral design data. The last two are considered and their possible improvements in accuracy by means of experimental measurements are assessed. The cost penalty on fuel cycle and capital cost due to each physical quantity is then considered, and consequently the benefits of reactor physics experiments are evaluated for a number of hypotheses concerning the foreseeable HTGR development and the delay in taking practical advantage of experimental results. It is concluded that, at the present state of knowledge of basic nuclear data and with the available calculation methods, the economic incentive to new reactor physics experiments is small, and a previous careful analysis is recommended to those intending to perform such experiments.

DABIE: a data banking system of integral experiments for reactor core characteristics computer codes

International Nuclear Information System (INIS)

Matsumoto, Kiyoshi; Naito, Yoshitaka; Ohkubo, Shuji; Aoyanagi, Hideo.

1987-05-01

A data banking system of integral experiments for reactor core characteristics computer codes, DABIE, has been developed to lighten the burden on searching so many documents to obtain experiment data required for verification of reactor core characteristics computer code. This data banking system, DABIE, has capabilities of systematic classification, registration and easy retrieval of experiment data. DABIE consists of data bank and supporting programs. Supporting programs are data registration program, data reference program and maintenance program. The system is designed so that user can easily register information of experiment systems including figures as well as geometry data and measured data or obtain those data through TSS terminal interactively. This manual describes the system structure, how-to-use and sample uses of this code system. (author)

Purification and solidification of reactor wastes at a Canadian nuclear generating station

International Nuclear Information System (INIS)

Buckley, L.P.; Burt, D.A.

1981-06-01

Chalk River Nuclear Laboratories are developing methods to condition power reactor wastes and to immobilize their radionuclides. Evaporation alone and combined with bituminization has been an important part of the program. After testing at the laboratories a 0.5 m 2 wiped-film evaporator was sent to the Douglas Point Nuclear Generating Station (220 MWe) to demonstrate its suitability to handle typical reactor liquid wastes. Two specific tasks undertaken with the wiped-film evaporator were successfully completed. The first was purification of contaminated heavy water which had leaked from the moderator circuit. The heavy water is normally recovered, cleaned by filters and ion-exchange resin and then upgraded by electrolysis. Cleaning the heavy water with the wiped-film evaporator produced better quality water for upgrading than had been achieved by any previous method and at much lower operating cost. The second task was to concentrate and immobilize a decontamination waste. The waste was generated from the decontamination of pump bowls used in the primary heat transport circuit. The simultaneous addition of the liquid waste and bitumen emulsion to the wiped-film evaporator produced a solid containing 30 wt% waste solids in a bitumen matrix. The volume reduction achieved was 16:1 based on the volumes of initial liquid waste and the final product generated. The quantity sent to storage was 20 times less than had the waste been immobilized in a cement matrix. The successful demonstration has resulted in a proposal to install a wiped-film evaporator at the station to clean heavy water and immobilize decontamination wastes. (author)

Heat Fluxes and Evaporation Measurements by Multi-Function Heat Pulse Probe: a Laboratory Experiment

Science.gov (United States)

Sharma, V.; Ciocca, F.; Hopmans, J. W.; Kamai, T.; Lunati, I.; Parlange, M. B.

2012-04-01

Multi Functional Heat Pulse Probes (MFHPP) are multi-needles probes developed in the last years able to measure temperature, thermal properties such as thermal diffusivity and volumetric heat capacity, from which soil moisture is directly retrieved, and electric conductivity (through a Wenner array). They allow the simultaneous measurement of coupled heat, water and solute transport in porous media, then. The use of only one instrument to estimate different quantities in the same volume and almost at the same time significantly reduces the need to interpolate different measurement types in space and time, increasing the ability to study the interdependencies characterizing the coupled transports, especially of water and heat, and water and solute. A three steps laboratory experiment is realized at EPFL to investigate the effectiveness and reliability of the MFHPP responses in a loamy soil from Conthey, Switzerland. In the first step specific calibration curves of volumetric heat capacity and thermal conductivity as function of known volumetric water content are obtained placing the MFHPP in small samplers filled with the soil homogeneously packed at different saturation degrees. The results are compared with literature values. In the second stage the ability of the MFHPP to measure heat fluxes is tested within a homemade thermally insulated calibration box and results are matched with those by two self-calibrating Heatflux plates (from Huxseflux), placed in the same box. In the last step the MFHPP are used to estimate the cumulative subsurface evaporation inside a small column (30 centimeters height per 8 centimeters inner diameter), placed on a scale, filled with the same loamy soil (homogeneously packed and then saturated) and equipped with a vertical array of four MFHPP inserted close to the surface. The subsurface evaporation is calculated from the difference between the net sensible heat and the net heat storage in the volume scanned by the probes, and the

Experimental Investigation Evaporation of Liquid Mixture Droplets during Depressurization into Air Stream

Science.gov (United States)

Liu, L.; Bi, Q. C.; Terekhov, Victor I.; Shishkin, Nikolay E.

2010-03-01

The objective of this study is to develop experimental method to study the evaporation process of liquid mixture droplets during depressurization and into air stream. During the experiment, a droplet was suspended on a thermocouple; an infrared thermal imager was used to measure the droplet surface temperature transition. Saltwater droplets were used to investigate the evaporation process during depressurization, and volatile liquid mixtures of ethanol, methanol and acetone in water were applied to experimentally research the evaporation into air stream. According to the results, the composition and concentration has a complex influence on the evaporation rate and the temperature transition. With an increase in the share of more volatile component, the evaporation rate increases. While, a higher salt concentration in water results in a lower evaporation rate. The shape variation of saltwater droplet also depends on the mass concentration in solution, whether it is higher or lower than the eutectic point (22.4%). The results provide important insight into the complex heat and mass transfer of liquid mixture during evaporation.

Mass hierarchy sensitivity of medium baseline reactor neutrino experiments with multiple detectors

Directory of Open Access Journals (Sweden)

Hong-Xin Wang

2017-05-01

Full Text Available We report the neutrino mass hierarchy (MH determination of medium baseline reactor neutrino experiments with multiple detectors, where the sensitivity of measuring the MH can be significantly improved by adding a near detector. Then the impact of the baseline and target mass of the near detector on the combined MH sensitivity has been studied thoroughly. The optimal selections of the baseline and target mass of the near detector are ∼12.5 km and ∼4 kton respectively for a far detector with the target mass of 20 kton and the baseline of 52.5 km. As typical examples of future medium baseline reactor neutrino experiments, the optimal location and target mass of the near detector are selected for the specific configurations of JUNO and RENO-50. Finally, we discuss distinct effects of the reactor antineutrino energy spectrum uncertainty for setups of a single detector and double detectors, which indicate that the spectrum uncertainty can be well constrained in the presence of the near detector.

Status of the Daya Bay Reactor Neutrino Oscillation Experiment

International Nuclear Information System (INIS)

Lin, Cheng-Ju Stephen

2010-01-01

The last unknown neutrino mixing angle θ 13 is one of the fundamental parameters of nature; it is also a crucial parameter for determining the sensitivity of future long-baseline experiments aimed to study CP violation in the neutrino sector. Daya Bay is a reactor neutrino oscillation experiment designed to achieve a sensitivity on the value of sin 2 (2*θ 13 ) to better than 0.01 at 90% CL. The experiment consists of multiple identical detectors placed underground at different baselines to minimize systematic errors and suppress cosmogenic backgrounds. With the baseline design, the expected anti-neutrino signal at the far site is about 360 events per day and at each of the near sites is about 1500 events per day. An overview and current status of the experiment will be presented.

Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

International Nuclear Information System (INIS)

Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

1983-02-01

This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

Emergency cooling system for nuclear reactors

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown

Observed and modeled multi-year evaporation from three field-scale experiments using water balance and Penman-Monteith methods: Profound effect of material type and wind exposure

Science.gov (United States)

Peterson, H. E.; Fretz, N.; Bay, D.; Mayer, K. U.; Smith, L.; Beckie, R. D.

2013-12-01

Three instrumented experimental waste-rock piles at the Cu-Zn-Mo Antamina Mine in Peru are composed of distinct types of waste rock but are otherwise almost identical in size and geometry and experience the same atmospheric conditions with the exception of wind exposure. Evaporation from the piles was calculated using the water balance method over three- and four-year periods to determine the effect of material type and meteorological variability on evaporation. Annual changes in water storage were low or negligible except as a result of unusually high annual precipitation. Observed evaporation was high (44% - 75% of precipitation) and was extremely variable annually in the coarsest-grained waste-rock pile 1, most likely as a result of greater wind exposure and air circulation in that pile. Observed evaporation was moderate (36% - 48% of precipitation) with moderate annual variability in the finer-grained, relatively homogeneous waste-rock pile 2. Observed evaporation was low (24% - 32% of precipitation) with low annual variability in the finer-grained, relatively heterogeneous waste-rock pile 3, most likely as a result of low air circulation coupled with complex flow regimes that include high-velocity preferential flow paths. Slightly higher evaporation was observed on the slopes than on the crowns of Pile 2, while much lower evaporation was observed on the slopes than on the crowns of Piles 1 and 3. Evidence suggests that Piles 1 and 3 slope water-balance evaporation estimates are skewed by non-vertical flow and that, in general, evaporation is higher on the slopes than on the crowns of the piles. Evaporation was also estimated using the Food and Agriculture Organization of the United Nations modified Penman-Monteith method (FAO-PM; Allen et al., 1998) using base-case laboratory- and software- derived parameters. The base-case method underestimated observed evaporation calculated by the water balance method for Pile 1, overestimated observed evaporation for Pile

Decommissioning of evaporation technology for processing liquid radioactive waste in UJV Rez, a. s

International Nuclear Information System (INIS)

Tous, M.; Otcovsky, T.; Podlaha, J.

2015-01-01

The UJV Rez, a. s. is the main leader in processing institutional radioactive waste (RAW) in the Czech Republic and the Waste Management Department has been established since the research reactor VVR-S (now LVR-15) was put in operation. Due to the large activities in nuclear research and engineering in the past, a big capacity of waste management technologies was needed. The low pressure compactor for volume reduction of solid RAW, as well as chemical pre-treatment technology of liquid RAW were installed and later the evaporation technology for effective processing the liquid RAW with the cementation and bituminization unit for final conditioning of concentrated liquid RAW were used. During the years of research reactor operation and research activities in UJV Rez, a. s. there were two installed evaporation technologies in row. After the latest evaporator lifetime, changes in liquid RAW production and together with higher decontamination factor requirements, this technology was decided to be decommissioned. The decommissioned evaporation technology was installed and put in operation in 1991. This technology was used for processing liquid aqueous RAW produced from internal research activities and of course for external producers and institutions (e.g. universities, medicine, research institutes, industry). The approved decommissioning plan was prepared and the licence for immediate decommissioning was obtained in 2012. Then the decommissioning project started. The preparing stages as dosimetric survey, expected material balance and of course initial decontamination activities were performed. Evaporation technology dismantling and processing the arising RAW were done by the internal staff of Waste Management Department. The total volume of produced RAW was 49,5 m 3 of RAW. The secondary liquid RAW (from decontamination) of amount 1,4 m 3 , contaminated sludge of amount 0,5 m 3 , solid RAW (construction steel) of amount 39,1 m 3 , solid compressible RAW (protective

Evaporation in hydrology and meteorology

NARCIS (Netherlands)

Brandsma, T.

1990-01-01

In this paper the role of evaporation in hydrology and meteorology is discussed, with the emphasis on hydrology. The basic theory of evaporation is given and methods to determine evaporation are presented. Some applications of evaporation studies in literature are given in order to illustrate the

Purification and solidification of reactor wastes at a Canadian nuclear generating station

International Nuclear Information System (INIS)

Buckley, L.P.; Burt, D.A.

1981-01-01

The study aimed at development and demonstration of volume reduction and solidification of CANDU reactor wastes has been underway at Chalk River Nuclear Laboratories in the Province of Ontario, Canada. The study comprises membrane separation processes, evaporator appraisal and immobilization of concentrated wastes in bitumen. This paper discusses the development work with a wiped-film evaporator and the successful completion of demonstration tests at Douglas Point Nuclear Generating Station. Heavy water from the moderator system was purified and wastes arising from pump bowl decontamination were immobilized in bitumen with the wiped-film evaporator that was used in the development tests at Chalk River

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

International Nuclear Information System (INIS)

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

1986-01-01

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

Review of operational experience with the gas-cooled Magnox reactors of the United Kingdom Central Electricity Generating Board

International Nuclear Information System (INIS)

Cave, L.; Clarke, A.W.

1984-01-01

The paper provides a review, which is mainly of a statistical nature, of 260 reactor years of operating experience which the (United Kingdom) Central Electricity Generating Board (CEGB) has obtained with its gas-cooled, graphite moderated Magnox reactors. The main emphasis in the review is on safety rather than on availability. Data are provided on the overall incidence and frequencies of faults and it is shown that the plant items which are predominantly responsible for recorded faults are the gas circulators and the turbo-alternators. Analysis of the reactor trip experience shows that the incidence of events which necessitate an automatic shutdown of the reactor has been about one per reactor year and that of other events leading to a reactor trip has not been much higher (1.4 per reactor year). As would be expected from the length of the operating experience, some relatively rare events have occurred (expected frequency 10 -2 per reactor year, or less) but on each occasion the reactor shutdown system and decay heat removal systems functioned satisfactorily. No overheating of, or damage to, the fuel occurred as a result of these rare events or of other, more frequent, faults. Analysis of the trend of failure rates has shown an improvement with time in nearly all safety-related items and external inspection of the primary coolant circuits has shown no significant deterioration with time. However, some derating of the reactors has been necessary to reduce the effects of oxidation of mild steel in CO 2 , in order to obtain optimum service lives. In spite of major differences between the systems, a comparison of the failure rates of analogous systems and plant items in PWRs and the Magnox reactors show a considerable similarity. Overall, the review of CEGB's operational experience with its Magnos reactors has shown that the frequencies of faults in systems and plant items has been satisfyingly low. (author)

Critical plasma-materials issues for fusion reactor designs

International Nuclear Information System (INIS)

Wilson, K.L.; Bauer, W.

1983-01-01

Plasma-materials interactions are a dominant driving force in the design of fusion power reactors. This paper presents a summary of plasma-materials interactions research. Emphasis is placed on critical aspects related to reactor design. Particular issues to be addressed are plasma edge characterization, hydrogen recycle, impurity introduction, and coating development. Typical wall fluxes in operating magnetically confined devices are summarized. Recent calculations of tritium inventory and first wall permeation, based on laboratory measurements of hydrogen recycling, are given for various reactor operating scenarios. Impurity introduction/wall erosion mechanisms considered include sputtering, chemical erosion, and evaporation (melting). Finally, the advanced material development for in-vessel components is discussed. (author)

Irradiation Experiments on Plutonium Fuels for Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Frost, B. R.T.; Wait, E. [Atomic Energy Research Establishment Harwell, Berks. (United Kingdom)

1967-09-15

An assessment carried out some years ago indicated that cermet fuels might provide the high burn-up and integrity required for fast reactors. An irradiation programme was started at Harwell on (U, Pu)O{sub 2} -SS cermet plates and rods, mainly In thermal neutron fluxes, to gain experience of dimensional stability at temperatures typical of modern sodium-cooled fast reactor designs (600-650 Degree-Sign C). A subsequent assessment showed that cermets carried a large penalty as far as breeding was concerned and (U, Pu)C was chosen by Harwell for long-term study as an alternative, economic, fast reactor fuel. However, the results from the cermet experiments were of sufficient promise to proceed with parallel irradiation programmes on cermets and carbide. The studies of cermets showed that dimensional instability (swelling and cladding rupture) were caused by the pressures exerted on the steel matrix by the fuel particles, and that the initial density of the fuel particles was important in determining the burn-up at which failure occurred. Further, it was shown that cermets provided a useful vehicle for studying the changes occurring in oxide fuel particles with increasing burn-up. The disappearance of initial porosity and its replacement by fission gas bubbles and segregated solid fission products was studied in some detaiL No significant differences were observed between UO{sub 2} and(U,Pu)O{sub 2} particles. The initial studies of (U, Pu)C were concerned with the effect of varying composition and structure on swelling and fission gas release. A tantalum-lined nickel alloy cladding material was used to contain both pellet and powder specimens In an irradiation experiment in the core of the Dounreay fast reactor. This showed that the presence of a metal phase in the fuel led to a high swelling rate, that fission gas release was low up to {approx} 3% bum-up, and that a low density powder accommodated the swelling without excessive straining of the can. A subsequent

Experience in using a research reactor for the training of power reactor operators

International Nuclear Information System (INIS)

Blotcky, A.J.; Arsenaut, L.J.

1972-01-01

A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Advanced Reactor Licensing: Experience with Digital I and C Technology in Evolutionary Plants

International Nuclear Information System (INIS)

Wood, RT

2004-01-01

This report presents the findings from a study of experience with digital instrumentation and controls (I and C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l and C systems and identified lessons learned. The report (1) gives an overview of the modern l and C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States

NEA Activities in Preserving, Evaluating and Applying Data from Fast Reactor Experiments

International Nuclear Information System (INIS)

Gulliford, N.T.; Cornet, S.M.; Hill, I.; Yamaji, A.

2015-01-01

The goal of the OECD Nuclear Energy Agency (NEA) in the area of nuclear science is to help member countries identify, collate, develop and disseminate the basic scientific and technical knowledge required to ensure safe and reliable operation of current nuclear systems and to develop next generation technologies. Within these general goals, the current nuclear science programme has three key objectives: (i) to help advance the existing scientific knowledge needed to enhance the performance and safety of current nuclear systems, (ii) to contribute to building a solid scientific and technical basis for the development of future generation nuclear systems and (iii) to support the preservation of essential knowledge in the field of nuclear science. As part of the second and third of these objectives, an extensive programme of work to preserve and evaluate data from integral experiments has been established, including reactor physics, shielding and criticality safety experiments on fast reactor systems. Data from experimental facilities are reviewed and, if necessary, archives of information are made safe. This may typically involve the indexing and scanning of key documents and archiving of logbooks, for example. Selected experiments go through a detailed evaluation process and where deemed appropriate, a benchmark description is created in a standardized format for inclusion in one of the NEA Data Bank international databases. This information is used extensively by the international nuclear science community to validate their modelling and simulation tools. The process can be viewed as part of a broader knowledge management function, where information is gathered, evaluated, linked and made accessible to a wide range of users. The presentation gives details of the main databases maintained and developed by the NEA, focusing on those related to fast reactor applications. The status of recent preservation activities for fast reactor archives in the United Kingdom is

Computer network that assists in the planning, execution and evaluation of in-reactor experiments

International Nuclear Information System (INIS)

Bauer, T.H.; Froehle, P.H.; August, C.; Baldwin, R.D.; Johanson, E.W.; Kraimer, M.R.; Simms, R.; Klickman, A.E.

1985-01-01

For over 20 years complex, in-reactor experiments have been performed at Argonne National Laboratory (ANL) to investigate the performance of nuclear reactor fuel and to support the development of large computer codes that address questions of reactor safety in full-scale plants. Not only are computer codes an important end-product of the research, but computer analysis is also involved intimately at most stages of experiment planning, data reduction, and evaluation. For instance, many experiments are of sufficiently long duration or, if they are of brief duration, occur in such a purposeful sequence that need for speedy availability of on-line data is paramount. This is made possible most efficiently by computer assisted displays and evaluation. A purposeful linking of main-frame, mini, and micro computers has been effected over the past eight years which greatly enhances the speed with which experimental data are reduced to useful forms and applied to the relevant technological issues. This greater efficiency in data management led also to improvements in the planning and execution of subsequent experiments. Raw data from experiments performed at INEL is stored directly on disk and tape with the aid of minicomputers. Either during or shortly after an experiment, data may be transferred, via a direct link, to the Illinois offices of ANL where the data base is stored on a minicomputer system. This Idaho-to-Illinois link has both enhanced experiment performance and allowed rapid dissemination of results

Surfactant-Enhanced Benard Convection on an Evaporating Drop

Science.gov (United States)

Nguyen, Van X.; Stebe, Kathleen J.

2001-11-01

Surfactant effects on an evaporating drop are studied experimentally. Using a fluorescent probe, the distribution and surface phase of the surfactant is directly imaged throughout the evaporation process. From these experiments, we identify conditions in which surfactants promote surface tension-driven Benard instabilities in aqueous systems. The drops under study contain finely divided particles, which act as tracers in the flow, and form well-defined patterns after the drop evaporates. Two flow fields have been reported in this system. The first occurs because the contact line becomes pinned by solid particles at the contact line region. In order for the contact line to remain fixed, an outward flow toward the ring results, driving further accumulation at the contact ring. A ‘coffee ring’ of particles is left as residue after the drop evaporates[1]. The second flow is Benard convection, driven by surface tension gradients on the drop[2,3]. In our experiments, an insoluble monolayer of pentadecanoic acid is spread at the interface of a pendant drop. The surface tension is recorded, and the drop is deposited on a well-defined solid substrate. Fluorescent images of the surface phase of the surfactant are recorded as the drop evaporates. The surfactant monolayer assumes a variety of surface states as a function of the area per molecule at the interface: surface gaseous, surface liquid expanded, and surface liquid condensed phases[4]. Depending upon the surface state of the surfactant as the drop evaporates, transitions of residue patterns left by the particles occur, from the coffee ring pattern to Benard cells to irregular patterns, suggesting a strong resistance to outward flow are observed. The occurrence of Benard cells on a surfactant-rich interface occurs when the interface is in LE-LC coexistence. Prior research concerning surfactant effects on this instability predict that surfactants are strongly stabilizing[5]. The mechanisms for this change in behavior

Thermal-Hydraulic Experiment To Test The Stable Operation Of A PIUS Type Reactor

International Nuclear Information System (INIS)

Irianto, Djoko; Kanji, T.; Kukita, Y.

1996-01-01

An advanced type of reaktor concept as the Process Inherent Ultimate Safety (PIUS) reactor was based on intrinsically passive safety considerations. The stable operation of a PIUS type reactor is based on the automation of circulation pump speed. An automatic circulation pump speed control system by using a measurement of the temperature distribution in the lower density lock is proposed the PIUS-type reactor. In principle this control system maintains the fluid temperature at the axial center of the lower density lock at average of the fluid temperatures below and above the lower density lock. This control system will prevent the poison water from penetrating into the core during normal operation. The effectiveness of this control system was successfully confirmed by a series of experiments using atmospheric pressure thermal-hydraulic test loop which simulated the PIUS principle. The experiments such as: start-up and power ramping tests for normal operation simulation and loss of feedwater test for an accident condition simulation, carried out in JAERI

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

Energy Technology Data Exchange (ETDEWEB)

T. A. Tomberlin; S. B. Grover

2004-11-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor

International Nuclear Information System (INIS)

T. A. Tomberlin; S. B. Grover

2004-01-01

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment

IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility

International Nuclear Information System (INIS)

Dietze, Klaus; Klippel, Henk Th.; Koning, Arjan; Jacqmin, Robert

2003-01-01

1 - Description: The STEK-experiments have been performed to check neutron data of the most important reactor materials, especially of fission product nuclides, fuel isotopes and structural materials. The measured central reactivity worths (CRW) of small samples were compared with calculated values. These C/E-ratios have been used then for data corrections or in adjustment procedures. The reactors STEK (ECN Petten/ Netherlands) was a fast-thermal coupled facility of zero power. The annular thermal drivers were filled by fuel assemblies and moderated by water. The inner insertion lattices were loaded with pellets of fuel and other materials producing the fast neutron flux. The characteristics of the neutron and adjoint spectra were obtained by special arrangements of these pellets in unit cells. In this way, a hard or soft neutron spectrum or a special energy behavior of the adjoint function could be reached. The samples were moved by means of tubes to the central position (pile-oscillation technique). The original information about the facility and measurements is compiled in RCN-209, ECN-10 The 5 STEK configurations cover a broad energy range due to their increasing softness. The experiments are very valuable because of the extensive program of sample reactivity measurements with many fission product nuclides important in reactor burn-up calculations. At first, analyses of the experiments have been performed in Petten. Newer analyses were done later in Cadarache / CEA France using the European scheme for reactor calculation JEF-2.2 / ECCO / ERANOS (see Note Techniques and JEF/DOC-746). Furthermore, re-analyses were performed in O-arai / JNC Japan with the JNC standard route JENDL-3.2 / SLAROM / CITATION / PERKY. Results obtained with both code systems and different data evaluations (JEF-2.2 and JENDL-3.2) are compared in JEF/DOC-861. It contains the following documents: 31 Reports, 2 publications, 5 JEF documents, 4 conferences. 2 - Related or auxiliary programs

Experimental Investigation of Double Effect Evaporative Cooling Unit

Directory of Open Access Journals (Sweden)

Ahmed Abd Mohammad Saleh

2018-03-01

Full Text Available This work presents the experimental investigation of double effect evaporative cooling unit with approximate capacity 7 kW. The unit consisted of two stages, the sensible heat exchanger and the cooling tower composing the external indirect regenerative evaporative cooling stage where a direct evaporative cooler represent the second stage. Testing results showed a maximum capacity and lowest supplied air temperature when the water flow rate in heat exchanger was 0.1 L/s. The experiment recorded the unit daily readings at two airflow rates (0.425 m3/s, 0.48 m3/s. The reading shows that unit inlet DBT is effect positively on unit wet bulb effectiveness and unit COP at constant humidity ratio. The air extraction ratio effected positively on the unit wet bulb effectiveness within a certain limit where maximum COP recorded 11.4 when the extraction ratio equal to 40%.

ALPHA visual data collection. STX005-025: melt drop steam explosion experiments

International Nuclear Information System (INIS)

Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun

1999-03-01

Steam explosion is a phenomenon in which a high temperature liquid gives its internal energy to a low temperature volatile liquid extremely quickly causing rapid evaporation and shock wave generation. In the field of nuclear reactor safety research regarding severe accidents in LWRs, steam explosions involving molten fuel and coolant has been recognized as a potential threat to the integrity of the reactor containment vessel. In the ALPHA (Assessment of Loads and Performance of Containment in Hypothetical Accident) program, experiments were performed to investigate the phenomenology of vapor explosions using iron-alumina thermite melt as a simulant of molten core. This report collects the experimental results especially emphasizing the visual observations by high speed photography. (author)

Experiment study on thermal mixing performance of HTR-PM reactor outlet

Energy Technology Data Exchange (ETDEWEB)

Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

2016-09-15

A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of the test facility is introduced, which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main mixing structure, hot gas duct, exhaust pipe system and I&C system. Experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed, which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

14th Biennial conference on reactor operating experience plant operations: The human element

International Nuclear Information System (INIS)

Anon.

1989-01-01

Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

Advanced Reactor Fuels Irradiation Experiment Design Objectives

Energy Technology Data Exchange (ETDEWEB)

Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

Advanced Reactor Fuels Irradiation Experiment Design Objectives

International Nuclear Information System (INIS)

Chichester, Heather Jean MacLean; Hayes, Steven Lowe; Dempsey, Douglas; Harp, Jason Michael

2016-01-01

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

Turkish Undergraduates' Misconceptions of Evaporation, Evaporation Rate, and Vapour Pressure

Science.gov (United States)

Canpolat, Nurtac

2006-01-01

This study focused on students' misconceptions related to evaporation, evaporation rate, and vapour pressure. Open-ended diagnostic questions were used with 107 undergraduates in the Primary Science Teacher Training Department in a state university in Turkey. In addition, 14 students from that sample were interviewed to clarify their written…

The reactor kinetics code tank: a validation against selected SPERT-1b experiments

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

Mode switching control of dual-evaporator air-conditioning systems

International Nuclear Information System (INIS)

Lin, J.-L.; Yeh, T.-J.

2009-01-01

Modern air-conditioners incorporate variable-speed compressors and variable-opening expansion valves as the actuators for improving cooling performance and energy efficiency. These actuators have to be properly feedback-controlled; otherwise the systems may exhibit even poorer performance than the conventional machines which use fixed-speed compressors and mechanical expansion valves. Particularly for an air-conditioner with multiple evaporators, there are occasions that the machine is operated in a mode that only selected evaporator(s) is(are) turned on, and switching(s) between modes occurs(occur) during the control process. In this case, one needs to have more carefully designed control and switching strategies to ensure the system performance. In this paper, a framework for mode switching control of the dual-evaporator air-conditioning (DEAC) system is proposed. The framework is basically an integration of a controller and a dynamic compensator. The controller, which possesses the flow-distribution capability and assumes both evaporators are on throughout the control process, is intended to provide nominal performance. While mode switching is achieved by varying the reference settings in the controller, the dynamic compensator is used to improve the transient responses immediately after the switching. Experiments indicate that the proposed framework can achieve satisfactory indoor temperature regulation and provide bumpless switching between different modes of operation.

Marangoni Convection in Evaporating Organic Liquid Droplets on a Nonwetting Substrate.

Science.gov (United States)

Chandramohan, Aditya; Dash, Susmita; Weibel, Justin A; Chen, Xuemei; Garimella, Suresh V

2016-05-17

We quantitatively characterize the flow field inside organic liquid droplets evaporating on a nonwetting substrate. A mushroom-structured surface yields the desired nonwetting behavior with methanol droplets, while use of a cooled substrate (5-15 °C) slows the rate of evaporation to allow quasi-static particle image velocimetry. Visualization reveals a toroidal vortex within the droplet that is characteristic of surface tension-driven flow; we demonstrate by means of a scaling analysis that this recirculating flow is Marangoni convection. The velocities in the droplet are on the order of 10-45 mm/s. Thus, unlike in the case of evaporation on wetting substrates where Marangoni convection can be ignored for the purpose of estimating the evaporation rate, advection due to the surface tension-driven flow plays a dominant role in the heat transfer within an evaporating droplet on a nonwetting substrate because of the large height-to-radius aspect ratio of the droplet. We formulate a reduced-order model that includes advective transport within the droplet for prediction of organic liquid droplet evaporation on a nonwetting substrate and confirm that the predicted temperature differential across the height of the droplet matches experiments.

Investigation of evaporation characteristics of polonium and its lighter homologues selenium and tellurium from liquid Pb-Bi-eutecticum

CERN Document Server

Neuhausen, J; Eichler, B

2004-01-01

The evaporation behaviour of polonium and its lighter homologues selenium and tellurium dissolved in liquid Pb-Bi-eutecticum (LBE) has been studied at various temperatures in the range from 482 K up to 1330 K under Ar/H2 and Ar/H2O-atmospheres using γ-ray spectroscopy. Polonium release in the temperature range of interest for technical applications is slow. Within short term (1h) experiments measurable amounts of polonium are evaporated only at temperatures above 973 K. Long term experiments reveal that a slow evaporation of polonium occurs at temperatures around 873 K resulting in a fractional polonium loss of the melt around 1% per day. Evaporation rates of selenium and tellurium are smaller than those of polonium. The presence of H2O does not enhance the evaporation within the error limits of our experiments. The thermodynamics and possible reaction pathways involved in polonium release from LBE are discussed.

GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; John D. Bess; Jim Gulliford

2011-09-01

Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

Background determination for the neutron-neutron scattering experiment at the reactor YAGUAR

Energy Technology Data Exchange (ETDEWEB)

Muzichka, A.Yu. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Furman, W.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Krylov, A.R. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E.I. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V.N. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A.V. [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Levakov, B.G. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Lyzhin, A.E. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Chernukhin, Yu.I. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Kandiev, Ya.Z. [Russian Federal Nuclear Center-All-Russian Research Institute of Technical Physics, PO Box 245, 456770 Snezhinsk (Russian Federation); Howell, C.R. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G.E. [North Carolina State University, Raleigh, NC 27695-8202 (United States); Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Crawford, B.E. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S.L. [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States)]. E-mail: sstephen@gettysburg.edu; Tornow, W. [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States)

2007-06-01

The motivation and design is outlined for the experiment to measure the neutron-neutron singlet scattering length directly with thermal neutrons at the pulsed reactor YAGUAR. A statistical accuracy of 3% can be reached, though achieving the goal of an overall accuracy of 3-5% for the nn-scattering length depends on the background level. Possible sources of background are discussed in depth and the results of extensive modeling of the background are presented. Measurements performed at YAGUAR to test these background calculations are described. The experimental results indicate an anticipated background level up to 30% relative to the expected nn effect at the maximal energy burst of the reactor. The conclusion is made that the nn experiment at YAGUAR is feasible to produce the first directly measured value for the neutron-neutron scattering length.

Comparative Analysis of Carbon Plasma in Arc and RF Reactors

International Nuclear Information System (INIS)

Todorovic-Markovic, B.; Markovic, Z.; Mohai, I.; Szepvolgyi, J.

2004-01-01

Results on studies of molecular spectra emitted in the initial stages of fullerene formation during the processing of graphite powder in induction RF reactor and evaporation of graphite electrodes in arc reactor are presented in this paper. It was found that C2 radicals were dominant molecular species in both plasmas. C2 radicals have an important role in the process of fullerene synthesis. The rotational-vibrational temperatures of C2 and CN species were calculated by fitting the experimental spectra to the simulated ones. The results of optical emission study of C2 radicals generated in carbon arc plasma have shown that rotational temperature of C2 species depends on carbon concentration and current intensity significantly. The optical emission study of induction RF plasma and SEM analysis of graphite powder before and after plasma treatment have shown that evaporation of the processed graphite powder depends on feed rate and composition of gas phase significantly. Based on the obtained results, it was concluded that in the plasma region CN radicals could be formed by the reaction of C2 species with atomic nitrogen at smaller loads. At larger feed rate of graphite powder, CN species were produced by surface reaction of the hot carbon particles with nitrogen atoms. The presence of nitrogen in induction RF plasma reduces the fullerene yield significantly. The fullerene yield obtained in two different reactors was: 13% in arc reactor and 4.1% in induction RF reactor. However, the fullerene production rate was higher in induction RF reactor-6.4 g/h versus 1.7 g/h in arc reactor

MHTGR: New production reactor summary of experience base

International Nuclear Information System (INIS)

1988-03-01

Worldwide interest in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) stems from the capability of the system to retain the advanced fuel and thermal performance while providing unparalleled levels of safety. The small power level of the MHTGR and its passive systems give it a margin of safety not attained by other concepts being developed for power generation. This report covers the experience base for the key nuclear system, components, and processes related to the MHTGR-NPR. 9 refs., 39 figs., 9 tabs

An assessment of the evaporation and condensation phenomena of lithium during the operation of a Li(d,xn fusion relevant neutron source

Directory of Open Access Journals (Sweden)

J. Knaster

2016-12-01

Full Text Available The flowing lithium target of a Li(d,xn fusion relevant neutron source must evacuate the deuteron beam power and generate in a stable manner a flux of neutrons with a broad peak at 14 MeV capable to cause similar phenomena as would undergo the structural materials of plasma facing components of a DEMO like reactors. Whereas the physics of the beam-target interaction are understood and the stability of the lithium screen flowing at the nominal conditions of IFMIF (25 mm thick screen with +/–1 mm surface amplitudes flowing at 15 m/s and 523 K has been demonstrated, a conclusive assessment of the evaporation and condensation of lithium during operation was missing. First attempts to determine evaporation rates started by Hertz in 1882 and have since been subject of continuous efforts driven by its practical importance; however intense surface evaporation is essentially a non-equilibrium process with its inherent theoretical difficulties. Hertz-Knudsen-Langmuir (HKL equation with Schrage’s ‘accommodation factor’ η = 1.66 provide excellent agreement with experiments for weak evaporation under certain conditions, which are present during a Li(d,xn facility operation. An assessment of the impact under the known operational conditions for IFMIF (574 K and 10−3Pa on the free surface, with the sticking probability of 1 inherent to a hot lithium gas contained in room temperature steel walls, is carried out. An explanation of the main physical concepts to adequately place needed assumptions is included.

Development of a lab-scale contaminated organic effluents treatment process using evaporation and supercritical water oxidation

Energy Technology Data Exchange (ETDEWEB)

Turc, H.A.; Joussot-Dubien, C

2004-07-01

The organic liquid waste produced in the ATALANTE facility have to be treated in order to reduce the fire and contamination risks. Therefore, the Mini-DELOS process has been developed, which combines a low pressure evaporator in a shielded enclosure and a continuous supercritical water oxidation (SCWO) reactor in a glovebox. Evaporation makes it possible to evacuate the main organic stream as decontaminated distillates to an industrial incinerator. The remaining residue, concentrating the radioactivity can be converted through SCWO into a contaminated aqueous effluent, fully compatible with the existing outlets of the facility. The preliminary results of the first year of active operation of the Mini- DELOS process are here presented. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-03-01

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

Chemistry in water reactors: operating experience and new developments. 2 volumes

International Nuclear Information System (INIS)

1994-01-01

These proceedings of the International conference on chemistry in water reactors (Operating experience and new developments), Volume 1, are divided into 8 sessions bearing on: (session 1) Primary coolant activity, corrosion products (5 conferences), (session 2) Dose reduction (4 conferences), (session 3) New developments (4 conferences), poster session: Primary coolant chemistry (16 posters), (session 4) Decontamination (5 conferences), poster session (2 posters), (session 5) BWR-Operating experience (3 conferences), (session 6) BWR-Modelling of operating experience (4 conferences), (session 7) BWR-Basic studies (4 conferences), (session 8) BWR-New technologies (3 conferences)

Recent performance experience with US light water reactor self-actuating safety and relief valves

Energy Technology Data Exchange (ETDEWEB)

Hammer, C.G.

1996-12-01

Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

Hydrothermal waves in evaporating sessile drops

OpenAIRE

Brutin, D.; Rigollet, F.; Niliot, C. Le

2009-01-01

Drop evaporation is a simple phenomena but still unclear concerning the mechanisms of evaporation. A common agreement of the scientific community based on experimental and numerical work evidences that most of the evaporation occurs at the triple line. However, the rate of evaporation is still empirically predicted due to the lack of knowledge on the convection cells which develop inside the drop under evaporation. The evaporation of sessile drop is more complicated than it appears due to the...

Experience of developing the imitators of the fuel element for the WWER reactors

International Nuclear Information System (INIS)

Balashov, S.M.; Boltenko, Eh.A.; Vinogradov, V.A.

1998-01-01

Peculiarities of designs of fuel elements imitators for the WWER-type reactors of nominal capacity and with single-ended current feed positioning are considered. The data on the filler heat conductivity and the results of tests and application of the fuel elements imitators at various testing facilities are presented. The possibility of equipping one of the non operating WWER reactors with the fuel element imitators for conduct of large-scale experiment is indicated

Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

International Nuclear Information System (INIS)

Halvers, L.J.

1977-03-01

An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Bitumen and cement solidifications of LL and ML liquid radwaste. The SGN experience

International Nuclear Information System (INIS)

Tchemitcheff, E.; Roux, P.

1993-01-01

The presentation is focused on the thin-film evaporator technology and the experience gained in the field of the NPPs and research centers on radwaste conditioning. As early as 1970, SGN was licensed by the CEA for the bituminization of LL and ML radwaste. With the support of EDF and COGEMA, SGN has been performing in depth research on cement solidification of borated concentrates and ion exchange resins generated by reactors or reprocessing plant since 1983

Melting and evaporation analysis of the first wall in a water-cooled breeding blanket module under vertical displacement event by using the MARS code

Energy Technology Data Exchange (ETDEWEB)

Kim, Geon-Woo [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148 Gwahak-ro, Yuseong-gu, Daejeon 34133 (Korea, Republic of)

2017-05-15

Highlights: • Material phase change of first wall was simulated for vertical displacement event. • An in-house first wall module was developed to simulate melting and evaporation. • Effective heat capacity method and evaporation model were proposed. • MARS code was proposed to predict two-phase phenomena in coolant channel. • Phase change simulation was performed by coupling MARS and in-house module. - Abstract: Plasma facing components of tokamak reactors such as ITER or the Korean fusion demonstration reactor (K-DEMO) can be subjected to damage by plasma instabilities. Plasma disruptions like vertical displacement event (VDE) with high heat flux, can cause melting and vaporization of plasma facing materials and burnout of coolant channels. In this study, to simulate melting and vaporization of the first wall in a water-cooled breeding blanket under VDE, one-dimensional heat equations were solved numerically by using an in-house first wall module, including phase change models, effective heat capacity method, and evaporation model. For thermal-hydraulics, the in-house first wall analysis module was coupled with the nuclear reactor safety analysis code, MARS, to take advantage of its prediction capability for two-phase flow and critical heat flux (CHF) occurrence. The first wall was proposed for simulation according to the conceptual design of the K-DEMO, and the heat flux of plasma disruption with a value of 600 MW/m{sup 2} for 0.1 s was applied. The phase change simulation results were analyzed in terms of the melting and evaporation thicknesses and the occurrence of CHF. The thermal integrity of the blanket first wall is discussed to confirm whether the structural material melts for the given conditions.

Does evaporation paradox exist in China?

Directory of Open Access Journals (Sweden)

Z. T. Cong

2009-03-01

Full Text Available One expected consequence of global warming is the increase in evaporation. However, lots of observations show that the rate of evaporation from open pans of water has been steadily decreasing all over the world in the past 50 years. The contrast between expectation and observation is called "evaporation paradox". Based on data from 317 weather stations in China from 1956 to 2005, the trends of pan evaporation and air temperature were obtained and evaporation paradox was analyzed. The conclusions include: (1 From 1956 to 2005, pan evaporation paradox existed in China as a whole while pan evaporation kept decreasing and air temperature became warmer and warmer, but it does not apply to Northeast and Southeast China; (2 From 1956 to 1985, pan evaporation paradox existed narrowly as a whole with unobvious climate warming trend, but it does not apply to Northeast China; (3 From 1986 to 2005, in the past 20 years, pan evaporation paradox did not exist for the whole period while pan evaporation kept increasing, although it existed in South China. Furthermore, the trend of other weather factors including sunshine duration, windspeed, humidity and vapor pressure deficit, and their relations with pan evaporation are discussed. As a result, it can be concluded that pan evaporation decreasing is caused by the decreasing in radiation and wind speed before 1985 and pan evaporation increasing is caused by the decreasing in vapor pressure deficit due to strong warming after 1986. With the Budyko curve, it can be concluded that the actual evaporation decreased in the former 30 years and increased in the latter 20 year for the whole China.

Evaporation-driven clustering of microscale pillars and lamellae

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae-Hong; Kim, Jungchul; Kim, Ho-Young, E-mail: hyk@snu.ac.kr [Department of Mechanical and Aerospace Engineering, Seoul National University, Seoul 08826 (Korea, Republic of)

2016-02-15

As a liquid film covering an array of micro- or nanoscale pillars or lamellae evaporates, its meniscus pulls the elastic patterns together because of capillary effects, leading to clustering of the slender microstructures. While this elastocapillary coalescence may imply various useful applications, it is detrimental to a semiconductor manufacturing process called the spin drying, where a liquid film rinses patterned wafers until drying. To understand the transient mechanism underlying such self-organization during and after liquid evaporation, we visualize the clustering dynamics of polymer micropatterns. Our visualization experiments reveal that the patterns clumped during liquid evaporation can be re-separated when completely dried in some cases. This restoration behavior is explained by considering adhesion energy of the patterns as well as capillary forces, which leads to a regime map to predict whether permanent stiction would occur. This work does not only extend our understanding of micropattern stiction, but also suggests a novel path to control and prevent pattern clustering.

Processes influencing cooling of reactor effluents

International Nuclear Information System (INIS)

Magoulas, V.E.; Murphy, C.E. Jr.

1982-01-01

Discharge of heated reactor cooling water from SRP reactors to the Savannah River is through sections of stream channels into the Savannah River Swamp and from the swamp into the river. Significant cooling of the reactor effluents takes place in both the streams and swamp. The majority of the cooling is through processes taking place at the surface of the water. The major means of heat dissipation are convective transfer of heat to the air, latent heat transfer through evaporation and radiative transfer of infrared radiation. A model was developed which incorporates the effects of these processes on stream and swamp cooling of reactor effluents. The model was used to simulate the effect of modifications in the stream environment on the temperature of water flowing into the river. Environmental effects simulated were the effect of changing radiant heat load, the effect of changes in tree canopy density in the swamp, the effect of total removal of trees from the swamp, and the effect of diverting the heated water from L reactor from Steel Creek to Pen Branch. 6 references, 7 figures

Microwave heating type evaporator

International Nuclear Information System (INIS)

Taura, Masazumi; Nishi, Akio; Morimoto, Takashi; Izumi, Jun; Tamura, Kazuo; Morooka, Akihiko.

1987-01-01

Purpose: To prevent evaporization stills against corrosion due to radioactive liquid wastes. Constitution: Microwaves are supplied from a microwave generator by way of a wave guide tube and through a microwave permeation window to the inside of an evaporatization still. A matching device is attached to the wave guide tube for transmitting the microwaves in order to match the impedance. When the microwaves are supplied to the inside of the evaporization still, radioactive liquid wastes supplied from a liquid feed port by way of a spray tower to the inside of the evaporization still is heated and evaporated by the induction heating of the microwaves. (Seki, T.)

Dynamic Roughness Ratio-Based Framework for Modeling Mixed Mode of Droplet Evaporation.

Science.gov (United States)

Gunjan, Madhu Ranjan; Raj, Rishi

2017-07-18

The spatiotemporal evolution of an evaporating sessile droplet and its effect on lifetime is crucial to various disciplines of science and technology. Although experimental investigations suggest three distinct modes through which a droplet evaporates, namely, the constant contact radius (CCR), the constant contact angle (CCA), and the mixed, only the CCR and the CCA modes have been modeled reasonably. Here we use experiments with water droplets on flat and micropillared silicon substrates to characterize the mixed mode. We visualize that a perfect CCA mode after the initial CCR mode is an idealization on a flat silicon substrate, and the receding contact line undergoes intermittent but recurring pinning (CCR mode) as it encounters fresh contaminants on the surface. The resulting increase in roughness lowers the contact angle of the droplet during these intermittent CCR modes until the next depinning event, followed by the CCA mode of evaporation. The airborne contaminants in our experiments are mostly loosely adhered to the surface and travel along with the receding contact line. The resulting gradual increase in the apparent roughness and hence the extent of CCR mode over CCA mode forces appreciable decrease in the contact angle observed during the mixed mode of evaporation. Unlike loosely adhered airborne contaminants on flat samples, micropillars act as fixed roughness features. The apparent roughness fluctuates about the mean value as the contact line recedes between pillars. Evaporation on these surfaces exhibits stick-jump motion with a short-duration mixed mode toward the end when the droplet size becomes comparable to the pillar spacing. We incorporate this dynamic roughness into a classical evaporation model to accurately predict the droplet evolution throughout the three modes, for both flat and micropillared silicon surfaces. We believe that this framework can also be extended to model the evaporation of nanofluids and the coffee-ring effect, among

Vapor-based interferometric measurement of local evaporation rate and interfacial temperature of evaporating droplets.

Science.gov (United States)

Dehaeck, Sam; Rednikov, Alexey; Colinet, Pierre

2014-03-04

The local evaporation rate and interfacial temperature are two quintessential characteristics for the study of evaporating droplets. Here, it is shown how one can extract these quantities by measuring the vapor concentration field around the droplet with digital holographic interferometry. As a concrete example, an evaporating freely receding pending droplet of 3M Novec HFE-7000 is analyzed at ambient conditions. The measured vapor cloud is shown to deviate significantly from a pure-diffusion regime calculation, but it compares favorably to a new boundary-layer theory accounting for a buoyancy-induced convection in the gas and the influence upon it of a thermal Marangoni flow. By integration of the measured local evaporation rate over the interface, the global evaporation rate is obtained and validated by a side-view measurement of the droplet shape. Advective effects are found to boost the global evaporation rate by a factor of 4 as compared to the diffusion-limited theory.

Reactor design, cold-model experiment and CFD modeling for chemical looping combustion

Energy Technology Data Exchange (ETDEWEB)

Zhang, Shaohua; Ma, Jinchen; Hu, Xintao; Zhao, Haibo; Wang, Baowen; Zheng, Chuguang [Huazhong Univ. of Science and Technology, Wuhan (China). State Key Lab. of Coal Combustion

2013-07-01

Chemical looping combustion (CLC) is an efficient, clean and cheap technology for CO{sub 2} capture, and an interconnected fluidized bed is more appropriate solution for CLC. This paper aims to design a reactor system for CLC, carry out cold-model experiment of the system, and model fuel reactor using commercial CFD software. As for the CLC system, the air reactor (AR) is designed as a fast fluidized bed while the fuel reactor (FR) is a bubbling bed; a cyclone is used for solid separation of the AR exit flow. The AR and FR are separated by two U-type loop seals to remain gas sealed. Considered the chemical kinetics of oxygen carrier, fluid dynamics, pressure balance and mass balance of the system simultaneously, some key design parameters of a CH{sub 4}-fueled and Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3}-based CLC reactor (thermal power of 50 kWth) are determined, including key geometric parameters (reactor cross-sectional area and reactor height) and operation parameters (bed material quantity, solid circulation rate, apparent gas velocity of each reactor). A cold-model bench having same geometric parameters with its prototype is built up to study the effects of various operation conditions (including gas velocity in the reactors and loop seals, and bed material height, etc.) on the solids circulation rate, gas leakage, and pressure balance. It is witnessed the cold-model system is able to meet special requirements for CLC system such as gas sealing between AR and FR, the circulation rate and particles residence time. Furthermore, the thermal FR reactor with oxygen carrier of Fe{sub 2}O{sub 3}/Al{sub 2}O{sub 3} and fuel of CH{sub 4} is simulated by commercial CFD solver FLUENT. It is found that for the design case the combustion efficiency of CH{sub 4} reaches 88.2%. A few part of methane is unburned due to fast, large bubbles rising through the reactor.

Thermogravimetric analysis of fuel film evaporation

Institute of Scientific and Technical Information of China (English)

HU Zongjie; LI Liguang; YU Shui

2006-01-01

Thermogravimetric analysis (TGA) was compared with the petrochemical distillation measurement method to better understand the characteristics of fuel film evaporation at different wall tem- peratures. The film evaporation characteristics of 90# gasoline, 93# gasoline and 0# diesel with different initial thicknesses were investigated at different environmental fluxes and heating rates. The influences of heating rate, film thickness and environmental flux on fuel film evaporation for these fuels were found. The results showed that the environmental conditions in TGA were similar to those for fuel films in the internal combustion engines, so data from TGA were suitable for the analysis of fuel film evaporation. TGA could simulate the key influencing factors for fuel film evaporation and could investigate the basic quantificational effect of heating rate and film thickness. To get a rapid and sufficient fuel film evaporation, sufficiently high wall temperature is necessary. Evaporation time decreases at a high heating rate and thin film thickness, and intense gas flow is important to promoting fuel film evaporation. Data from TGA at a heating rate of 100℃/min are fit to analyze the diesel film evaporation during cold-start and warming-up. Due to the tense molecular interactions, the evaporation sequence could not be strictly divided according to the boiling points of each component for multicomponent dissolved mixture during the quick evaporation process, and the heavier components could vaporize before reaching their boiling points. The 0# diesel film would fully evaporate when the wall temperature is beyond 250℃.

Water-evaporation reduction by duplex films: application to the human tear film.

Science.gov (United States)

Cerretani, Colin F; Ho, Nghia H; Radke, C J

2013-09-01

Water-evaporation reduction by duplex-oil films is especially important to understand the physiology of the human tear film. Secreted lipids, called meibum, form a duplex film that coats the aqueous tear film and purportedly reduces tear evaporation. Lipid-layer deficiency is correlated with the occurrence of dry-eye disease; however, in-vitro experiments fail to show water-evaporation reduction by tear-lipid duplex films. We review the available literature on water-evaporation reduction by duplex-oil films and outline the theoretical underpinnings of spreading and evaporation kinetics that govern behavior of these systems. A dissolution-diffusion model unifies the data reported in the literature and identifies dewetting of duplex films into lenses as a key challenge to obtaining significant evaporation reduction. We develop an improved apparatus for measuring evaporation reduction by duplex-oil films including simultaneous assessment of film coverage, stability, and temperature, all under controlled external mass transfer. New data reported in this study fit into the larger body of work conducted on water-evaporation reduction by duplex-oil films. Duplex-oil films of oxidized mineral oil/mucin (MOx/BSM), human meibum (HM), and bovine meibum (BM) reduce water evaporation by a dissolution-diffusion mechanism, as confirmed by agreement between measurement and theory. The water permeability of oxidized-mineral-oil duplex films agrees with those reported in the literature, after correction for the presence of mucin. We find that duplex-oil films of bovine and human meibum at physiologic temperature reduce water evaporation only 6-8% for a 100-nm film thickness pertinent to the human tear film. Comparison to in-vivo human tear-evaporation measurements is inconclusive because evaporation from a clean-water surface is not measured and because the mass-transfer resistance is not characterized. Copyright © 2013 Elsevier B.V. All rights reserved.

A study of the evaporation of heterogeneous water droplets under active heating

Science.gov (United States)

Piskunov, Maxim; Legros, Jean Claude; Strizhak, Pavel

2016-11-01

Using high-speed video registration tools with a sample rate of 102-104 frames per second (fps), we studied the patterns in the evaporation of water droplets containing 1 and 2 mm individual metallic inclusions in a high-temperature gas environment. The materials of choice for the inclusions were steels (AISI 1080 carbon steel and AISI type 316L stainless steel) and pure nickel. We established the lifetimes τh of the liquid droplets under study with a controlled increase in the gas environment temperature up to 900 K. We also considered the physical aspects behind the τh distribution in the experiments conducted and specified the conditions for more effective cooling of metallic inclusions. Following the experimental research findings, a method was devised for effective reactor vessel cooling to avoid a meltdown at a nuclear power plant. The optimization of heat and mass transfer modes was performed within the framework of the strategic plan for the development of National Research Tomsk Polytechnic University as one of the world-leading universities.

Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

International Nuclear Information System (INIS)

Bagley, K.Q.

1978-01-01

The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

Nuclear engineering laboratory self regulated power oscillation experiments at the Health Physics Research Reactor

International Nuclear Information System (INIS)

Miller, L.F.; Mihalczo, J.T.; Bailiff, E.G.; Woody, N.D.; Gardner, G.D.

1983-01-01

Self regulated power oscillation experiments with a variety of initial conditions have been performed with the ORNL Health Physics Research Reactor (HPRR) by undergraduate nuclear engineering students from The University of Tennessee for several years. These experiments demonstrate the coupling between reactor kinetics and heat transfer and show how the temperature coefficient of reactivity affects reactor behavior. A model that consists of several coupled first order nonlinear differential equations is used to calculate the temperature of the core center and surface and power as a function of time which are compared with the experimental data; also, the model is also used to study the effects of various model parameters and initial conditions on the amplitude, frequency and damping of the power and temperature oscillations. A previous paper presented some limited experimental results and demonstrated the correspondence between a simple point model and the experimental data. This paper presents the results of experiments for: (1) the initial power fixed at 9 kW with central core temperatures of 300 0 F and 500 0 F, annd (2) the initial central core temperature fixed at 500 0 F with initial powers of 6 and 8 kW

Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

International Nuclear Information System (INIS)

Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

1987-09-01

Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work

Non-aqueous removal of sodium from reactor components

Energy Technology Data Exchange (ETDEWEB)

Welch, F H; Steele, O P [Rockwell International, Atomics International Division, Canoga Park (United States)

1978-08-01

Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component.

Non-aqueous removal of sodium from reactor components

International Nuclear Information System (INIS)

Welch, F.H.; Steele, O.P.

1978-01-01

Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component

Reactor Neutrino Detection for Non Proliferation with the NUCIFER Experiment

Energy Technology Data Exchange (ETDEWEB)

Bouvet, L. [CEA, Centre de Saclay, IRFU, Gif-sur-Yvette, (France); Bouvier, S.; Bui, V. M. [Laboratoire Subatech, Ecole des Mines, Nantes Cedex 3 (France); others, and

2012-06-15

Neutrinos are the most abundant matter particles in the Universe. Thoroughly investigated in basic science, the neutrino field is now delivering first applications to the monitoring of nuclear reactors. The neutrinos are emitted in the decay chain of the fission products; therefore measuring their flux provides real-time information, directly related to the fission process occurring in the reactor core. Because of the very weak interaction of neutrinos with matter a neutrino detector can stand outside the core containment vessel and provide a non-intrusive and inherently tamper resistant measurement. After a brief review of the existing data and worldwide projects, we present the NUCIFER experiment. The active part of the detector is a tank filled up with one ton of Gadolinium-doped liquid scintillator. Sixteen photomultiplier tubes, isolated from the liquid by an acrylic buffer, read out the light produced by the interaction of a neutrino with the protons of the liquid. The tank is surrounded by plastic scintillator plates to veto the cosmic rays. Then polyethylene and lead shielding suppress the background coming from external neutrons and gamma rays respectively. The NUCIFER detector has been designed for an optimal compromise between the detection performances and the specifications of operation in a safeguards regime. Its global footprint is 2.8 m x 2.8 m and it can monitor remotely the nuclear power plant thermal power and Plutonium content with very little maintenance on years scale. The experiment is currently installed near the OSIRIS research reactor (70 MWth) at CEA, in Saclay, France. First data are expected by May 2012. This work is done in contact with the IAEA/SGTN division that is currently investigating the potentiality of neutrinos as a novel safeguards tool. A dedicated working group has been created in 2010 to coordinate the simulation effort of various reactor types as well as the development of dedicated detectors and define and eventually