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Sample records for natural reactor oklo

  1. Oklo natural reactor

    International Nuclear Information System (INIS)

    Fujii, Isao

    1985-01-01

    In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

  2. The Oklo natural nuclear reactors in Gabon

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After a recall of the first experiments of fission chain reaction within the first man-made nuclear reactor, the author describes how the formation of the Earth resulted in the presence of radioactive isotopes, and recalls how the existence of the natural reactor was discovered in the 1970's: measurements revealed a content of uranium hexafluoride which was abnormally but only slightly smaller than normal. The author gives some explanations presently given to the Oklo phenomenon, and wanders whether Oklo is a natural analogue of geological storage

  3. Oklo natural reactors: geological and geochemical conditions

    International Nuclear Information System (INIS)

    Jakubick, A.T.; Church, W.

    1986-02-01

    Published as well as unpublished material on the Oklo natural reactors in Gabon was evaluated with regard to the long-term aspects of nuclear waste disposal. Even though the vast data base available at present can provide only a site specific description of the phenomenon, already this material gives relevant information on plutonium retention, metamictization, fission product release, hydrogeochemical stability and migration of fission products. Generalized conclusions applicable to other nuclear waste repository would require the quantitative reconstruction of t s coupled thermo-hydrologic-chemical processes. This could be achieved by studying the deviations in the 2 H/ 1 H and 18 O/ 16 O ratios of minerals at Oklo. A further generalization of the findings from Oklo could be realized by examining the newly-discovered reactor zone 10, which was active under very different thermal conditions than the other reactors. 205 refs

  4. The nonlinear dynamics of the Oklo natural reactor

    International Nuclear Information System (INIS)

    Bilanovic, Z.; Harms, A.A.

    1985-01-01

    An analysis of the Oklo natural reactor, a self-sustaining and self-regulating critical assembly that existed some 2 billion years ago in Gabon, Africa, is presented. Nonlinear continuous dif ferential and nonlinear discrete iterative formulations are established and selected parameter characterizations identified. Conceivable power oscillations are calculated and discussed. Some implications of nonlinear mappings for nuclear simulation are suggested

  5. The Oklo reactors

    International Nuclear Information System (INIS)

    Skytte Jensen, B.

    1982-01-01

    The Oklo reactors comprise up to nine 235-U depleted zones in an uranium ore in the Republic of Gabon in West Africa. The depletion in fissile U-235 has been proved to have caused by nuclear chain reactions. The study of the Oklo phenomenon indicates that very efficient retardation mechanisms may operate in nature - at least under special conditions. A closer study of these processes ought to be made to establish the limitations to their occurrence. The Oklo sandstone formation today would probably be considered unacceptable as a host rock for a repository. (EG)

  6. The deposit of Oklo and its natural nuclear reactors

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Weber, F.; Pfiffelmann, J.P.; Chauvet, R.; Michel, B.; Reboul, J.C.

    1980-01-01

    In the uranium deposit of Oklo (Republic of Gabon), seven zones have been discovered since 1972, in which natural fission reactions took place. Since 1974, a thorough geological study of these zones has been undertaken. It includes field studies, observations of drilled samples and laboratory studies. These studies permit the authors to define the geological environment of the reactors and to point out the influence of nuclear reactions on the surrounding formations. All this work was completed by a geological and metallogenic study of the deposit of Oklo and of the uraniferous basin of Franceville. The deposit of Oklo is situated in a detrital, sandstone-like and pelitic series belonging to the Francevillian. The Francevillian and the mineralization are dated as Middle Precambrian (1800-2000 M.A.). The ore of Oklo is the result of two concentration stages. In the first, uranium seems to have been fixed by hydrocarbons that were concentrated in oil traps. After a tectonic event, circulations of oxidizing solutions generated reconcentrations that are associated with hematite and have contents of UO 2 between 1 and 20%. The fission reactions developed in the high-graded ores which had formed during the last phase of UO 2 concentration. A thorough tectonic analysis of the ore deposit shows that high-graded ores and fission reactors are controlled by fractures. The working of nuclear reactors results in a local increase of temperature which gave a rise to circulation of warm water. The results of this hydrothermal circulation and of the neutron bombardment are seen in a succession of facies surrounding the reactors. At the centre of the reactor all sedimentary structures have been destroyed; within the reaction zone the following clays mineral zones are founded: (1) 1 Md illite and ferrous chlorite corresponding to the common Francevillian sediment; (2) 2 Md illite, (3) magnesium chlorite and (4) 1 Md illite and chlorite-vermiculite in the very rich uraninite ore

  7. Natural fission reactors - the Oklo phenomenon

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Overview describes the discovery of the site, location of the reactors and site geology and discusses the permanence of fission products, nuclear reaction control mechanisms and trace concentrations of elements that act as poisons. (Author)

  8. Oklo reactors: natural analogs to nuclear waste repositories

    International Nuclear Information System (INIS)

    Curtis, D.B.; Benjamin, T.M.; Gancarz, A.J.

    1981-01-01

    The 2-billion-year-old fossil reactors at Oklo are ancient natural nuclear waste sites. Isotope dilution mass spectrometric analyses of the fission products in the reactor core uraninite and the peripheral pelitic sandstone provide data for calculating the reactor operating parameters, the quantities of fissiogenic isotopes produced, the fraction of these isotopes retained in the cores, and the location in the peripheral rocks of the fissiogenic fraction lost from the cores. For a duration of criticality of 3 x 10 5 yrs, the thermal plus resonance neutron fluence ranged between 10 20 and 10 21 neutrons/cm 2 . The fraction of technetium (60 to 85%), ruthenium (75 to 90%), and neodymium (85 to 100%) retained is negatively correlated with fluence. The lost fission products are contained within a few tens of meters of their source, the reactor cores. The systematics of the decay of 99 Tc (t/sub 1/2/ = 2.13 x 10 5 yr) to 99 Ru limits the period of fissiogenic element migration to approximately 1 million yr at a time 2 billion yr ago. Thermodynamic calculations of the temperature-dependent solubilities indicate that the loss of fissiogenic elements is diffusion controlled, whereas retention in the surrounding rocks is a result of temperature-dependent deposition from an aqueous solution. These results concerning the geochemistry of technetium, ruthenium, and neodymium at a natural waste site support the concept of geologic burial of man-made radioactive wastes

  9. Radiolysis in nature: Evidence from the Oklo natural reactors

    International Nuclear Information System (INIS)

    Curtis, D.B.; Gancarz, A.J.

    1983-02-01

    An examination of the mineralogy of the reactor zones at Oklo shows that they have been significantly altered. The rocks immediately adjacent to these zones are also mineralogically modified with respect to normal uranium bearing rocks. The mineralogic changes appear to be the consequence of radiation damage, changes in the bulk chemistry of the system and increased temperatures. Chemical changes were the consequence of convectively circulating fluids that transported elements in and out of the rocks. There were also changes in the electrochemical conditions in the rocks. These changes can most reasonably be attributed to oxidizing and reducing species produced by the radiolysis of water. We have calculated radiation doses and examined the production of radiolysis products in the fluid phase which lead to the following conclusions: 1) There was a net reduction of iron, probably associated with a net increase in total iron in the rocks of the reactor zones. The reduction of iron was most likely the result of hydrogen produced by the radiolysis of water. 2) Commensurate with the iron reduction, there was an oxidation of uranium and multivalent fission products, resulting in their transport out of the reactor zone. 3) Approximately 10 percent of the uranium and various proportions of these fission products were removed and redeposited in rocks within a few meters of the reactor zones. 4) The calculated radiation doses from alpha radiation and the inferred hydrogen production suggest an effective radiation yield of 0.06 molecules of hydrogen per 100 eV of energy imparted to the fluid phase. Considering radiation from both alpha and beta sources, the G value for hydrogen production is reduced to 0.01 to 0.002 molecules H 2 /100 eV. (author)

  10. OKLO: fossil reactors

    International Nuclear Information System (INIS)

    Naudet, R.

    Events leading up to the discovery during the summer of 1972 of the Oklo fossil reactor in Gabon and its subsequent exploration are reviewed. Results of studies are summarized; future investigations are outlined

  11. Relevance of the studies of the OKLO natural nuclear reactors to the storage of radioactive wastes

    International Nuclear Information System (INIS)

    Hagemann, R.; Roth, E.

    1978-01-01

    The geological environment of the OKLO natural nuclear reactors is described along with the operating caracteristics of the reactors. Data relevant to the stability of most of the fission products and to the transuranium elements in the reaction zones are reviewed. (orig.) [de

  12. Fate of the Epsilon Phase in the Oklo Natural Reactors

    International Nuclear Information System (INIS)

    S. Utsunomiya; R.C. Ewing

    2005-01-01

    In spent nuclear fuel (SNF), the micron- to submicron-sized epsilon phase (Mo-Ru-Pd-Tc-Rh) is an important host of 99 Tc which has a long half life (2.13 x 10 5 years) and can be an important contributor to dose in safety assessments of nuclear waste repositories. In addition, Tc is predominantly present as TcO 4 - under oxidizing conditions at wide range of pH, weakly adsorbed onto mineral surfaces, and unlikely to be incorporated into alteration uranyl minerals. In the Oklo natural reactor (2.0 Ga), essentially all of the 99 Tc has decayed to 99 Ru. Thus, this study focuses on Ru and the other metals of the epsilon phase in order to investigate the occurrence and the fate of the epsilon phase during the corrosion of this natural SNF. Samples from reactor zone (RZ)-10 (836, 819, 687); from RZ-13 (864, 910); were investigated using TEM (transmission electron microscopy). Within the UO 2 matrix, a Bi-Pd particle (40-60 nm), fioodite, PdBi 2 , was observed with trace amounts of As, Fe, and Te surrounded by an amorphous Pb-rich area. (Pd,Rh) 2 As, palladodymite or rhodarsenide, was observed (400-500 nm in size). Ruthenarsinite, (Ru,Ni)As, was identified in most samples: with a representative composition of As, 59.9: Co, 2.5: Ni, 5.2; Ru, 18.6; Rh, 8.4; Pd, 3.1; Sb, 2.4 in atomic percent. The particles diameters are a few hundred nanometers and, in most cases, surrounded by a Pb-rich phase (400-500 nm). Typically, the ruthenarsenite does not occur as single particle but an aggregate of ∼200 nm-sized particles. Some Ru-particles revealed a complex phase separation within the grain such as a Ru-particle (600-700 nm) with Pb at the core of the particle and enrichment of Ni, Co, and As at the rim. Some ruthenarsenite crystals were embedded in chlorite immediately adjacent to uraninite. A few particles were still coated by Pb. These results suggest a history for the epsilon phases: (1) The original epsilon phase was transformed to, in most cases, ruthenarsenite. (2) All

  13. Geochemical properties and nuclear chemical characteristics of Oklo natural fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, Hiroshi [Hiroshima Univ., Higashi-Hiroshima (Japan). Faculty of Science

    1997-07-01

    There are six uranium deposits in the Gabonese Republic in the cnetral Africa. `Fission reactor zone`, the fission chain reactions generated about 200 billion years ago, was existed in a part of them. CEA begun geochemical researches of Oklo deposits etc. in 1991. The geochemical and nuclear chemical properties of Oklo were reviewed from the results of researches. Oklo deposits is consisted of main five sedimentary faces such as sandstone (FA), Black Shale formation (FB), mudstone (FC), tuff (FD) and volcaniclastic sandstone (FE) from the bottom on the base rock of granite in the Precambrian era. Uranium is enriched in the upper part of FA layer and the under part of FB layer. {sup 235}U/{sup 238}U, U content, fission proportion, duration time, neutron fluence, temperature, restitution factor of {sup 235}U and epithermal index ({gamma}) were investigated and compared. The geochemical properties of Oklo are as followed: large enrich of uranium, the abundance ratio of {sup 235}U as same as that of enriched uranium, interaction of natural water and small rear earth elements. These factors made casually Oklo fission reactor. (S.Y.)

  14. Role of organic matter in the Proterozoic Oklo natural fission reactors, Gabon, Africa

    International Nuclear Information System (INIS)

    Nagy, B.; Rigali, M.J.; Gauthier-Lafaye, F.; Holliger, P.; Mossman, D.J.; Leventhal, J.S.

    1993-01-01

    Of the sixteen known Oklo and the Bangombe natural fission reactors (hydrothermally altered elastic sedimentary rocks that contain abundant uraninite and authigenic clay minerals), reactors 1 to 6 at Oklo contain only traces of organic matter, but the others are rich in organic substances. Reactors 7 to 9 are the subjects of this study. These organic-rich reactors may serve as time-tested analogues for anthropogenic nuclear-waste containment strategies. Organic matter helped to concentrate quantities of uranium sufficient to initiate the nuclear chain reactions. Liquid bitumen was generated from organic matter by hydrothermal reactions during nuclear criticality. The bitumen soon became a solid, consisting of polycyclic aromatic hydrocarbons and an intimate mixture of cryptocrystalline graphite, which enclosed and immobilized uraninite and the fission-generated isotopes entrapped in uraninite. This mechanism prevented major loss of uranium and fission products from the natural nuclear reactors for 1.2 b.y. 24 refs., 4 figs

  15. Organic matter and containment of uranium and fissiogenic isotopes at the Oklo natural reactors

    International Nuclear Information System (INIS)

    Nagy, B.; Rigali, M.J.; Davis, D.W.; Parnell, J.

    1991-01-01

    Some of the Precambrian natural fission reactors at Oklo in Gabon contain abundant organic matter, part of which was liquefied at the time of criticality and subsequently converted to a graphitic solid. The liquid organic matter helps to reduce U(VI) to U(IV) from aqueous solutions, resulting in the precipitation of uraninite. It is known that in the prevailing reactor environments, precipitated uraninite grains incorporated fission products. We report here observations which show that these uraninite crystals were held immobile within the re-solidified, graphitic bituminous organics at Oklo thus enhanced radionuclide containment. Uraninite encased in solid graphitic matter in the organic-rich reactor zones lost virtually no fissiogenic lanthanide isotopes. The first major episode of uranium and lead migration was caused by the intrusion of a swarm of adjacent dolerite dykes about 1,100 Myr after the reactors went critical. Our results from Oklo imply that the use of organic, hydrophobic solids such as graphitic bitumen as a means of immobilizing radionuclides in pre-treated nuclear waste warrants further investigation. (author)

  16. Organic free radicals and micropores in solid graphitic carbonaceous matter at the Oklo natural fission reactors, Gabon

    International Nuclear Information System (INIS)

    Rigali, M.J.; Nagy, B.

    1997-01-01

    The presence, concentration, and distribution of organic free radicals as well as their association with specific surface areas and microporosities help characterize the evolution and behavior of the Oklo carbonaceous matter. Such information is necessary in order to evaluate uranium mineralization, liquid bitumen solidification, and radio nuclide containment at Oklo. In the Oklo ore deposits and natural fission reactors carbonaceous matter is often referred to as solid graphitic bitumen. The carbonaceous parts of the natural reactors may contain as much as 65.9% organic C by weight in heterogeneous distribution within the clay-rich matrix. The solid carbonaceous matter immobilized small uraninite crystals and some fission products enclosed in this uraninite and thereby facilitated radio nuclide containment in the reactors. Hence, the Oklo natural fission reactors are currently the subjects of detailed studies because they may be useful analogues to support performance assessment of radio nuclide containment at anthropogenic radioactive waste repository sites. Seven carbonaceous matter rich samples from the 1968 ± 50 Ma old natural fission reactors and the associated Oklo uranium ore deposit were studied by electron spin resonance (ESR) spectroscopy and by measurements of specific surface areas (BET method). Humic acid, fulvic acid, and fully crystalline graphite standards were also examined by ESR spectroscopy for comparison with the Oklo solid graphitic bitumens. With one exception, the ancient Oklo bitumens have higher organic free radical concentrations than the modem humic and fulvic acid samples. The presence of carbon free radicals in the graphite standard could not be determined due to the conductivity of this material. 72 refs., 7 figs., 1 tab

  17. The neutron balance of the natural reactors at Oklo

    International Nuclear Information System (INIS)

    Naudet, R.; Filip, A.

    1975-01-01

    The authors discuss the main parameters determining criticality: the concentration of fissionable nuclei in the uranium; concentration of neutron-capturing nuclei in the gangue; concentration of uranium in the ore and rearrangement of the uranium to form ''critical masses''; amount of water present. Moderation was caused partly by the water of constitution of the clays in the gangue. Examination of the available data indicates that criticality could quite well have been achieved. A computer code (BINOCLE) was written for handling the neutron physics problems raised by the natural reactors. This very simple code, which can nevertheless handle the important points in sufficient detail, is well suited for describing the ores, providing a clear breakdown of the neutron balance and the quantities necessary for interpreting the analyses. It is designed to serve as a subprogram to a series of other codes: one-dimensional criticality; point evolution; spatial evolution; consideration of thermal transfers. Results showing the role of the main parameters are presented. The physical quantities measured by fission-product analysis are also found: proportions of fast fissions; conversion coefficient; spectral indices

  18. 99Tc, Pb and Ru migration around the Oklo natural fission reactors

    International Nuclear Information System (INIS)

    Gancarz, A.; Cowan, G.; Curtis, D.; Maeck, W.

    1980-01-01

    This work demonstrates the utility of the Oklo uranium ore deposit and natural fission reactors as a long time scale analogue for man-made radioactive waste repositories. It has been shown that the ores and nearby rocks were open to the loss and gain of 99 Tc, ruthenium, and lead relative to uranium. Identified regions of element deficiencies and those which are correspondingly enriched are separated by less than 10 meters. However, more extensive sampling is required to define the overall extent of the element migration. Element fractionation took place on at least two vastly different time scales; 99 Tc was fractionated from ruthenium within one million years of the end of reactor criticality. Lead-uranium fractionation has been ongoing for most of the two billion years since the ores were formed. Diffusion loss of lead from host uraninite appears to be an important process in the fractionation of lead from uranium

  19. Oklo reactors and implications for nuclear science

    OpenAIRE

    Davis, E. D.; Gould, C. R.; Sharapov, E. I.

    2014-01-01

    We summarize the nuclear physics interests in the Oklo natural nuclear reactors, focusing particularly on developments over the past two decades. Modeling of the reactors has become increasingly sophisticated, employing Monte Carlo simulations with realistic geometries and materials that can generate both the thermal and epithermal fractions. The water content and the temperatures of the reactors have been uncertain parameters. We discuss recent work pointing to lower temperatures than earlie...

  20. The Oklo fossil reactor

    International Nuclear Information System (INIS)

    Roth, Etienne.

    1975-01-01

    From the observation of anomalous 235 U content of a UF 6 cylinder in Pierrelatte, it was possible to trace back this anomaly to minerals coming from the Oklo quarry in Gabon. Large variations in 235 U content such as were observed could only come from very specific processes, one of them being induced fission. To investigate this hypothesis it was looked for the fission rare earths and their isotopic composition, and these unequivocally assigned the phenomenon to a reaction of fission of 235 U [fr

  1. The Oklo reactor

    International Nuclear Information System (INIS)

    McNeil, Russell

    1986-01-01

    The construction of a reactor, capable of producing a controlled nuclear chain reaction, has been one of the most complex achievements of modern science. That a similar reaction might take place in nature did not play a role in the thinking of the nuclear scientists responsible for it. Yet, 14 years ago, French scientists discovered that just such a phenomenon apparently occurred in western Africa almost 2 billion years ago. In this article Russell McNeil describes this fascinating curiosity and a recent attempt to model it mathematically

  2. Natural nuclear reactor at Oklo and variation of fundamental constants: Computation of neutronics of a fresh core

    International Nuclear Information System (INIS)

    Petrov, Yu. V.; Nazarov, A. I.; Onegin, M. S.; Petrov, V. Yu.; Sakhnovsky, E. G.

    2006-01-01

    Using modern methods of reactor physics, we performed full-scale calculations of the Oklo natural reactor. For reliability, we used recent versions of two Monte Carlo codes: the Russian code MCU-REA and the well-known international code MCNP. Both codes produced similar results. We constructed a computer model of the Oklo reactor zone RZ2 which takes into account all details of design and composition. The calculations were performed for three fresh cores with different uranium contents. Multiplication factors, reactivities, and neutron fluxes were calculated. We also estimated the temperature and void effects for the fresh core. As would be expected, we found for the fresh core a significant difference between reactor and Maxwell spectra, which had been used before for averaging cross sections in the Oklo reactor. The averaged cross section of 62 149 Sm and its dependence on the shift of a resonance position E r (due to variation of fundamental constants) are significantly different from previous results. Contrary to the results of previous papers, we found no evidence of a change of the samarium cross section: a possible shift of the resonance energy is given by the limits -73≤ΔE r ≤62 meV. Following tradition, we have used formulas of Damour and Dyson to estimate the rate of change of the fine structure constant α. We obtain new, more accurate limits of -4x10 -17 ≤α·/α≤3x10 -17 yr -1 . Further improvement of the accuracy of the limits can be achieved by taking account of the core burn-up. These calculations are in progress

  3. Oklo natural fission reactor program. Progress report, April 1-August 31, 1980

    International Nuclear Information System (INIS)

    Curtis, D.B.

    1980-12-01

    An interim report has been published on the redistribution of uranium, thorium, and lead in samples representing several million cubic meters of sandstone and metamorphosed sediments in the Athabasca Basin which is located in the northwest corner of the Canadian province of Saskatchewan. The region of study includes zones of uranium mineralization at Key Lake. Mineralization occurs at the unconformity between the Athabasca sandstone and the underlying metasediments and in fault zones within the metasediments. Lead isotopes record a radiometric age of 1300 +- 150 m.y. in samples from above and below the unconformity. This age probably reflects the time of deposition of the sandstones and an associated redistribution of uranium and/or lead in the underlying rocks. Many of the samples have been fractionated with respect to radiogenic lead and the actinide parent elements since that time. Sandstones and altered rocks from the region above the unconformity have been a transport path and are a repository for lead. In contrast, mineralized rocks are deficient in radiogenic lead and must be an important source of lead in the local geologic environment. Samples from Oklo reactor zone 9 and nearby host rocks have been prepared for isotopic analyses of ruthenium, molybdenum, uranium and lead

  4. OKLO: Fossil nuclear reactors. Physical study

    International Nuclear Information System (INIS)

    Naudet, R.

    1991-04-01

    This book presents a study of Oklo reactors, based essentially on physics and particularly neutronics but reviewing also all what is known on this topic, regrouping observations, measurement results and interpretative calculations. A remarkable characteristic of the study is the use of sophisticated reactor calculation methods for analysis of what happened two billion years ago in a uranium deposit. 200 refs [fr

  5. The Oklo natural nuclear reactors: neutron parameters, age and duration of the reactions, uranium and fission products migrations

    International Nuclear Information System (INIS)

    Ruffenach, J.-C.

    1979-09-01

    Mass spectrometry and isotopic dilution technique are used in order to carry out, on various samples from the fossil nuclear reactors at Oklo, Gabon, isotopic and chemical analyses of some particular elements involved in the nuclear reactions: uranium, lead, bismuth, thorium, rare gases (krypton, xenon), rare earths (neodymium, samarium, europium, gadolinium, dysprosium), ruthenium and palladium. Interpretations of these analyses lead to the determination of many neutron parameters such as the neutron fluence received by the samples, the spectrum index, the conversion coefficient, and also the percentages of fissions due to uranium-238 and plutonium-239 and the total number of fissions relative to uranium. All these results make it possible to determine the age of the nuclear reactions by measuring the amounts of fission rare earths formed, i.e. 1.97 billion years. This study brings some informations to the general problem of radioactive wastes storage in deep geological formations, the storage of uranium, plutonium and many fission products having been carried out naturally, and for about two billion years [fr

  6. Uranium redistribution under oxidizing conditions in Oklo natural reactor zone 2, Gabon

    International Nuclear Information System (INIS)

    Isobe, H.; Ohnuki, T.; Murakami, T.; Gauthier-Lafaye, F.

    1995-01-01

    This mineralogical study was completed to elucidate the relationships between uranium distribution and alteration products of the host rock of natural reactor zone clays just below the reactor core. Uraninite is preserved without any alteration in the reactor core. Uranium minerals are found to be present in the fractures in the reactor zone clays associated with iron-mineral veins, galena and Ti-bearing minerals. Uranium, for which the phases could not be identified, occurs in iron-mineral veins and the iron-mineral rim of pyrite grains in the reactor zone clays. Uranium is not associated with granular iron minerals occurring in the illite matrix of the reactor zone clays. The degree of crystallinity and uranium content of the three iron-bearing alteration products suggest that they formed under different conditions; the granular iron minerals, under alteration conditions where uranium was not mobilized while the iron-mineral veins and the iron-mineral rim of pyrite, under conditions in which uranium is mobilized after the formation of the granular iron minerals

  7. Fluid phases contemporary with sandstone diagenesis tectonic movements and the activity of the natural nuclear reactors in the Oklo deposit, Republic of Gabon

    International Nuclear Information System (INIS)

    Poty, B.

    1979-12-01

    A research contract was agreed between the IAEA and the Centre de Recherches Petrographiques et Geochimiques of Nancy (France) to analyze some aspects of the fluid phases present during diagenesis of the Precambrian Francevillean sandstones of Gabon, which include the Oklo uranium deposit in which because of the geometry of the ore body, the high ore grade, etc., a natural uranium reactor was formed. The investigation was orientated to define some special characteristics of the fluid inclusions and two main methods were applied for this purpose: The Raman spectroscopy (MOLE microsonde) and microthermometric analysis. The main conclusions of the research are the following: 1. The Francevillean sandstones were buried up to a depth of 4-5 km where the corresponding geothermal temperature was of around 240 0 C but during the Oklo nuclear reaction, the fluid temperatures were higher than 450 0 C and in at least one case (Zone II) up to 600 0 C. 2. The tectonic fracturing has favoured the fluid circulation, which was possibly the responsible of the mineral re-concentration after the Oklo nuclear reaction. 3. The diagenetic fluids were essentially aqueous solutions and no sulphur components were identified. 4. The hydrogen presence in a quartz veinlet is surprising and possibly due to water decomposition by strong irradiation

  8. Eh-pH diagrams for elements from Z = 40 to Z = 52: application to the Oklo natural reactor, Gabon

    Energy Technology Data Exchange (ETDEWEB)

    Brookins, D G [New Mexico Univ., Albuquerque (USA). Dept. of Geology

    1978-11-01

    Eh-pH diagrams for elements from Z = 40 (Zr) to Z = 52 (Te) have been constructed in order to comment on migration/retention of these elements at Oklo. Although data for fissiogenic amounts of some of these elements are lacking, where such data are available to agreement between predicted migration/retention based on the Eh-pH diagrams and actual measurement is excellent. Based on Eh-pH diagrams, migration (to what degree is uncertain) of Mo and Cd is predicted whereas retention of Y, Zr, Tc, Ru, Rh, Pd, Ag and Te is also predicted. An earlier report of Frejacques et al. of Tc migration is in disagreement with Eh-pH prediction, and recent (unpublished) data argue for Tc retention. In view of the agreement between prediction and observation, the possible migration of Sb and retention of In and Sn is proposed. These data again demonstrate the usefulness of Eh-pH diagrams for the Oklo fossil nuclear reactor but, more important, allow constraints to be placed on repositories for nuclear waste now under consideration.

  9. Mineralogical and petrographic investigations at Strasbourg on the Oklo natural reactors

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Weber, F.

    1978-01-01

    Petrographic and mineralogical investigations have revealed the presence around the reaction zones of aureoles characterized by the nature of the phyllitic minerals which they enclose. Running from the outside towards the core, these are: illite 1Md and ferriferous chlorite (normal sediment), illite 2M 1 , magnesium chlorite and kaolinite, illite 1M and vermiculite chlorite. Detritic quartz dissolution figures are observed which disappear in the illite 2M aureole or in the magnesium chlorite aureole, depending on the degree of sandiness of the country rock. This zonal stucture could be attributable to the combined effect of neutron bombardment and hydrothermal alteration due to the action of a thermal syphon triggered off by the nuclear reactions. (author)

  10. Far field hydrogeochemistry in the Oklo reactor area (Gabon)

    International Nuclear Information System (INIS)

    Toulhoat, P.; Gallien, J.P.; L'Henoret, P.

    1993-01-01

    In the frame of a general study of the Oklo natural reactor, which takes into account the natural analogue aspect, a complete hydrogeological and hydrogeochemical study is undertaken. The partners of this study are the following: - Section de geochimie, CEA (France): P. Toulhoat, J.P. Gallien, P. L'Henoret, V. Moulin (groundwater chemistry and colloids). - Ecole des Mines de Paris (CIG, Fontainebleau) E. Ledoux, I. Gurban (hydrogeology and modelling) - SKB and Conterra AB (Sweden) J.A.T. Smellie, A. Winberg (hydrogeology, isotope geochemistry). The aim of this study is to try to understand and to characterize the possible mobilization of elements or isotopes when groundwaters come in contact with nuclear reaction zones. The first step of the study is presented here, which comprises a general geochemical and hydrodynamical characterization of the site. In this presentation, the site of Bagombe is also mentioned as it has been confirmed as sector in which nuclear fission reactions occurred as in Oklo. (author). 10 refs., 6 figs., 6 tabs

  11. Geological follow-up of the mining of the Oklo natural rectors

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.

    1978-01-01

    The part of the Oklo natural reactors which was still in place after their discovery was mined between September 1975 and February 1977. In the reaction zones 27 faces and 8 outcrops have been surveyed in detail and sampled, approximately 1700 samples having been taken. In the course of these surveys a large number of sedimentological, petrographic and tectonic observations have been made. (author)

  12. Gas, benefits and question marks. The Oklo reactors: 100 % natural. The Kyoto protocol: use it or lose it?. Small hydro power: a great leap forward. The energy mix of South Korea

    International Nuclear Information System (INIS)

    Anon.

    2005-01-01

    This issue of Alternatives newsletter contains a main press-kit about natural gas economics worldwide and 4 articles dealing with the Oklo natural reactor, the Kyoto protocol, the small hydro-power in China, and the energy mix of South Korea: 1 - 'Gas benefits and question marks': The world's most widely distributed fossil fuel, natural gas is also the fastest-growing energy source of the past thirty years. Its position as the fuel of choice in the global energy mix is due in large part to its many domestic and industrial applications. 2 - 'The Oklo reactors: 100% natural': Another look at this extraordinary 2 billion year-old phenomenon in words and pictures: the nuclear fission reaction that created the natural reactors of Gabon. 3 - 'The Kyoto Protocol: use it or lose it?': Nearly eight years after its signature, the Kyoto Protocol is still hotly debated. Two experts give us their views: Spencer Abraham, former U.S. Secretary for Energy, and Jean-Charles Hourcade of CIRED, the international center for research on the environment and development. 4 - 'Small hydro power: a great leap forward': The Chinese government has responded to the need for rural electrification with an aid program for the country's poorest cantons. Enter the small hydro plant in northern Guangxi province. 5 - 'The energy mix of South Korea': Faced with continuing strong economic growth and energy demand, South Korea has multiplied its projects, from hydropower to tidal power to nuclear and even hydrogen in the longer term

  13. Uranium deposits of Gabon and Oklo reactors. Metallogenic model for rich deposits of the lower proterozoic

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.

    1986-05-01

    The geology of the Franceville basin (Gabon) is examined: stratigraphy, tectonics and geodynamics. The mobile zone of the Ogooue is specially studied: lithology, metamorphism and tectonics, isotopic geochronologic data are given. The different uranium deposits are described. A whole chapter is devoted to the study of Oklo natural nuclear reactor. A metallogenic model is proposed evidencing conditions required for deposit genesis. Tectonics, microstructures sedimentology, organic matter, diagenesis and uraniferous mineralizations are examined [fr

  14. Oklo 2 Billion Years Before Fermi

    International Nuclear Information System (INIS)

    Barre, B.

    2005-02-01

    The author aims to present the little-known story of the Oklo natural reactors. He recalls the historical aspects of the Oklo reactors discovery by the CEA in 1972, he explains the scientific phenomenon and the interest, notably as a ''natural analogue'' for the geological disposal of high level radioactive wastes. (A.L.B.)

  15. Recent outputs of the Oklo (Gabon) natural analogue study to nuclear waste disposal

    International Nuclear Information System (INIS)

    Michaud, V.; Trotignon, L.; Louvat, D.

    2000-01-01

    In the past twenty five years, the natural nuclear reactors of Oklo have been the subject of numerous detailed studies. First investigated for the physical and neutron aspects of the nuclear reaction, they were then reconsidered because they provide a unique opportunity in the world to study the containment of actinides and fission products in a geological formation over a broad timescale (two billion years). Although the sites investigated do not represent a complete analogue of a repository system, many of the processes studied (mass transfer to the surface, transport, migration / retention), the spatial extent of these processes, and the timescales involved, are compatible with processes liable to occur during the lifespan of a repository for the deep geological disposal of spent nuclear fuel. A fresh program was therefore initiated as a European Commission project in 1990, entitled''Oklo as a natural analog for transfer processes in a radioactive waste repository'- phase 7, and then extended by a phase 2 entitled Oklo, Natural Analogue - Behavior of Nuclear Reaction Products in a Natural Environment''. Researches conducted in phase I served to determine the physical conditions of the operation of the natural reactor, reconstruct the geological history of the reactor environment, and decode the behavior of actinides as well as fission products in the surrounding geological formations. Phase N, which ended in June 1999, had three main objectives: i) to assess radionuclide migration and retention processes from the reactor zones to the geological environment, ii) to define the confinement properties and long-term behavior of geological materials; iii) to test models of processes related to radionuclide migration and retention, and eventually to provide suitable data and scenarios for performance assessment of nuclear waste disposal. This paper proposes a synthesis of the main outputs of the Oklo project to the performance assessment of nuclear waste disposal, the

  16. The Oklo reactors: five years of exploration of the site

    International Nuclear Information System (INIS)

    Naudet, R.

    1978-01-01

    The main phases of the exploration of the Oklo site since the discovery of the ''reactor'' phenomenon are outlined briefly. Over 180 sampling holes were drilled during the interruption of the mining activities in the sector concerned. Several new zones have been found. Mining was resumed in the second half of 1975, providing an opportunity for highly fruitful geological follow-up work: more precise knowledge was gained of the morphology of the reactors, and very many additional samples were taken. Plant treatment of the ore and the systematic analysis of batches have made it possible to establish a balance of missing uranium-235. A small portion containing sites of intense reaction has been preserved by being anchored to the quarry wall. Mining in this sector has now finished, but new indications of fission have been found, especially in the Okelobondo sector. (author)

  17. Geochemical behaviour study of radionuclides and their radiogenic daughters in the vicinity of Oklo 10 and 13 natural nuclear reactors (Gabon) - Application to high-level radioactive waste disposal

    International Nuclear Information System (INIS)

    Menet-Dressayre, Catherine

    1992-01-01

    Since 1981, the discovery of new and almost unaltered natural nuclear reactors in the uranium mine of Oklo (Gabon) renewed the interest of scientific community. Indeed, due to their specific features, these reactors could be extensively investigated as natural analogues to better understand the geochemical processes which may occur in a high level nuclear waste repository. The aim of this PhD thesis is to determine the present distribution of a few radionuclides or their radiogenic daughters initially formed within the reaction zones and to infer their geochemical behaviour, subsequently to the stopping of nuclear reactions. Our study was focused on reactors 10 and 13 and their immediate sandstone surroundings in order to decipher the fate of U, Y and light rare earth elements which are assumed to be chemical analogues of actinides and fission products. Mineralogical observations, chemical and isotopic analyses on bulk rocks, led us to conclude that a part of radionuclides, as well as their daughters, remained confined within the reactions zones, in association with secondary mineral phases, whereas another part migrated towards tbe reactor rims. The radionuclides were concentrated at the reactor border or migrated within the first few metres of the surrounding sandstone, according to the intensity of nuclear reactors and the presence of the so-called 'facies argile de pile' which constitutes an intermediate facies between that of reactor cores and that of the surrounding sandstone. In the latter, long range elemental transfers occurred via fissures. Some of them, contemporaneous to the nuclear reactions drained radionuclides-rich fluids at temperatures of about 150-170 deg. C. More recent fissures, observed only in the environment of reactor 13, have allowed the transport of hotter hydrothermal fluids (about 310 deg. C), likely related to the nearby intrusion of dolerite dyke. The principal implications of this work for the disposal of nuclear wastes

  18. Oklo: The fossil nuclear reactors. Physics study - Translation of chapters 6, 13 and conclusions

    Energy Technology Data Exchange (ETDEWEB)

    Naudet, R [CEA, Paris (France)

    1996-09-01

    Three parts of the 1991 book `Oklo: reacteurs nucleaires fossiles. Etude physique` have been translated in this report. The chapters bear the titles `Study of criticality`(45 p.), `Some problems with the overall functioning of the reactor zones`(45 p.) and `Conclusions` (15 p.), respectively.

  19. Oklo: The fossil nuclear reactors. Physics study - Translation of chapters 6, 13 and conclusions

    International Nuclear Information System (INIS)

    Naudet, R.

    1996-09-01

    Three parts of the 1991 book 'Oklo: reacteurs nucleaires fossiles. Etude physique' have been translated in this report. The chapters bear the titles 'Study of criticality'(45 p.), 'Some problems with the overall functioning of the reactor zones'(45 p.) and 'Conclusions' (15 p.), respectively

  20. Reconstitution of fluid paleo-circulations and element migrations in the environments of Oklo's natural nuclear reactors (Gabon) and of Tournemire's argillites (France)

    International Nuclear Information System (INIS)

    Mathieu, Regis

    1999-01-01

    To better characterize the mobilization and migration process in rocks, a petrological and geochemical study of fluid paleo-circulation through fractures has been made in two different sites: (1) The environment of natural nuclear reactors from Proterozoic Oklo uranium ores (Gabon). The Archean basement typical of TTG series and the sandstones-pelites series of the Franceville basin are affected by a fracturing mainly filled by quartz-daphnite-calcite-sulfides and barren ou mineralized bitumens. Three paragenetic stages has been correlated to three regional structural phases. During the first extensional phase, a low saline (1.7-6.5 wt% NaCl), heated in the basement (190-210 deg. C) and impoverished in 18 O meteoric recharge is injected into the basin, along major N-S faults. It was responsible of silicification. The circulation of diagenetic brines is able to leach U, Pb, Zr, REEs and P resulting from accessory minerals alteration, at the basin-scale between 2104 Ma and 1719 Ma (Pb/Pb isochrone obtained on galena incorporated in zircons). These brines are responsible of anomalous Th/La ratios (1.8) of FA silicified sandstones higher than those (0.25) of most of Archean and Proterozoic metasediments. They are highly chlorine, calco-sodic ([Cl] > 6 m, from 28 wt% NaCl to 30 wt% CaCl 2 ), equilibrated with carbonate and evaporitic layers of FA sandstones, with low temperatures (130 deg. C) and rich in Ca, Li and Br. They are expulsed laterally due to the compaction of FA sandstones, and upwards along sub-vertical fractures. During the second extensional phase, the mineralization stage, mainly controlled by N-S faults corresponds to a mixing (155-220 deg. C) between the brines, the meteoric recharge and hydrocarbons C 9 and C 10 -rich fluids derived from organic matter maturation in the FB pelites. The interaction of the three fluids is responsible of the mineralization in sandstones and in calcites displaying an organic carbon origin (δ 13 C=-10 to -15 0/00 vs. PDB

  1. Oklo 2 Billion Years Before Fermi; Les reacteurs naturels d'Oklo (Gabon): 2 milliards d'annees avant Fermi

    Energy Technology Data Exchange (ETDEWEB)

    Barre, B

    2005-02-15

    The author aims to present the little-known story of the Oklo natural reactors. He recalls the historical aspects of the Oklo reactors discovery by the CEA in 1972, he explains the scientific phenomenon and the interest, notably as a 'natural analogue' for the geological disposal of high level radioactive wastes. (A.L.B.)

  2. Radioactive wastes in Oklo

    International Nuclear Information System (INIS)

    Balcazar, M.; Flores R, J.H.; Pena, P.; Lopez, A.

    2006-01-01

    The acceptance of the Nuclear Energy as electric power supply implies to give answer to the population on the two main challenges to conquer in the public opinion: the nuclear accidents and the radioactive wastes. Several of the questions that are made on the radioactive wastes, its are the mobility migration of them, the geologic stability of the place where its are deposited and the possible migration toward the aquifer mantels. Since the half lives of the radioactive waste of a Nuclear Reactor are of several hundred of thousands of years, the technical explanations to the previous questions little convince to the public in general. In this work summary the results of the radioactive waste generated in a natural reactor, denominated Oklo effect that took place in Gabon, Africa, it makes several thousands of millions of years, a lot before the man appeared in the Earth. The identification of at least 17 reactors in Oklo it was carried out thanks to the difference in the concentrations of Uranium 235 and 238 prospective, and to the analysis of the non-mobility of the radioactive waste in the site. It was able by this way to determine that the reactors with sizes of hardly some decimeter and powers of around 100 kilowatts were operating in intermittent and spontaneous form for space of 150,000 years, with operation cycles of around 30 minutes. Recent studies have contributed information valuable on the natural confinement of the radioactive waste of the Oklo reactors in matrixes of minerals of aluminum phosphate that caught and immobilized them for thousands of millions of years. This extracted information from the nature contributes guides and it allows 'to verify' the validity of the current proposals on the immobilization of radioactive wastes of a nuclear reactor. This work presents in clear and accessible form to the public in general on the secure 'design', operation, 'decommissioning' and 'storage' of the radioactive waste of the reactors that the nature put

  3. Contribution to a summary of the results obtained in the field of geology on the Oklo natural reactors and their environment

    International Nuclear Information System (INIS)

    Weber, F.

    1978-01-01

    The very special characteristics of the ore in the reaction zones may be attributed to the effects of the nuclear reactions themselves: de-structuration of the gangue by neutronic effects and hydrothermal leaching due to the generation of a true thermal syphon by the nuclear reactions. This mechanism contributes to the propagation of reactions from a number of initial reaction sites. The latter were initiated in certain parts of the deposit, enriched as a result of a tectonic episode, by oxidizing currents in the shear troughs. These enriched zones occur as anomalies within a pre-existent deposit, the formation of which still poses many questions of a metallogenic nature. (author)

  4. Indications of uranium transport around the reactor zone at Bagombe (Oklo)

    International Nuclear Information System (INIS)

    Gurban, I.; Laaksoharju, M.; Ledoux, E.; Made, B.; Salignac, A.L.

    1998-08-01

    The aim of this study is to use the hydrogeological and hydrochemical data from Oklo Natural Analogue to compare the outcome of two independent modelling approaches (HYTEC-2D and M3) which can be used to model natural conditions surrounding the reactor. HYTEC-2D represents a 2D, deterministic, transport and multi-solutes reactive coupled code developed at Ecole des Mines de Paris. M3 (named Multivariate Mixing and Mass balance) is a mathematical-statistical concept code developed for SKB. The M3 results are visualised using the Voxel Analyst code and the outcome of the uranium transport predictions are made from a performance assessment point of view. This exercise was in the beginning intended to represent a validation for M3, by comparing this statistic approach with the standard hydrodynamic - geochemical coupled code HYTEC-2D. It was realized that the codes complete each other and a better understanding of the geochemical studied system is obtained. Thus, M3 can relatively easily be used to calculate mixing portions and to identify sinks or sources of element concentrations that may exist in a geochemical system. This can help to address the reactions in the coupled code such as HYTEC-2D, to identify the hydrodynamic and hydrochemical system and to reduce the computation time. M3 shows the existence of the buffer around the reactor. No transport of uranium was indicated downstream the reactor. HYTEC-2D gives the same result in the case when we consider the existence of the redox buffer in the model. M3 shows an increase of the alkalinity in the reactor zone. The increase of the alkalinity was indicated by the M3 modelling to be associated with microbial decomposition of organic material which added reducing capacity to the system. The modelling result was supported by new results from the last field campaign, which included in-situ Eh measurements and microbial sampling and identification. The effects from the same process was indicated also by the HYTEC-2D

  5. Uranium transport around the reactor zone at Okelobondo (Oklo). Data evaluation with M3 and HYTEC

    International Nuclear Information System (INIS)

    Gurban, I.; Laaksoharju, M.; Made, B.; Ledoux, E.

    1999-12-01

    The Swedish Nuclear Fuel and Waste Management Company (SKB) is conducting and participating in Natural Analogue activities as part of various studies regarding the final disposal of high level nuclear waste (HLW). The aim of this study is to use the hydrogeological and hydrochemical data from Okelobondo (Oklo Natural Analogue) to compare the outcome of two independent modelling approaches (HYTEC and M3). The modelling helps to evaluate the processes associated with nuclear natural reactors such as redox, adsorption/desorption and dissolution/precipitation of the uranium and to develop more realistic codes which can be used for site investigations and data evaluation. HYTEC (1D and 2D) represents a deterministic, transport and multi-solutes reactive coupled code developed at Ecole des Mines de Paris. M3 (Multivariate Mixing and Mass balance calculations) is a mathematical-statistical concept code developed for SKB. M3 can relatively easily be used to calculate mixing portions and to identify sinks or sources of element concentrations that may exist in a geochemical system. M3 helped to address the reactions in the coupled code HYTEC. Thus, the major flow-paths and reaction paths were identified and used for transport evaluation. The reactive transport results (one-dimensional and two-dimensional simulations) are in good agreement with the statistical approach using the M3 model. M3 and HYTEC show a dissolution of the uranium layer in contact with upwardly oxidising waters. M3 and HYTEC show a gain of manganese rich minerals downstream the reactor. A comparison of the U and Mn plots for M3 deviation and HYTEC results showed an almost mirror behaviour. The U transport stops when the Mn gain increases. Thus, HYTEC and M3 modelling predict that a possible reason for not having U transport up to the surface in Okelobondo is due to an inorganic trap which may hinder the uranium transport. The two independent modelling approaches can be used to complement each other and to

  6. Radioactive wastes in Oklo; Desechos radiactivos en Oklo

    Energy Technology Data Exchange (ETDEWEB)

    Balcazar, M.; Flores R, J.H.; Pena, P.; Lopez, A. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2006-07-01

    The acceptance of the Nuclear Energy as electric power supply implies to give answer to the population on the two main challenges to conquer in the public opinion: the nuclear accidents and the radioactive wastes. Several of the questions that are made on the radioactive wastes, its are the mobility migration of them, the geologic stability of the place where its are deposited and the possible migration toward the aquifer mantels. Since the half lives of the radioactive waste of a Nuclear Reactor are of several hundred of thousands of years, the technical explanations to the previous questions little convince to the public in general. In this work summary the results of the radioactive waste generated in a natural reactor, denominated Oklo effect that took place in Gabon, Africa, it makes several thousands of millions of years, a lot before the man appeared in the Earth. The identification of at least 17 reactors in Oklo it was carried out thanks to the difference in the concentrations of Uranium 235 and 238 prospective, and to the analysis of the non-mobility of the radioactive waste in the site. It was able by this way to determine that the reactors with sizes of hardly some decimeter and powers of around 100 kilowatts were operating in intermittent and spontaneous form for space of 150,000 years, with operation cycles of around 30 minutes. Recent studies have contributed information valuable on the natural confinement of the radioactive waste of the Oklo reactors in matrixes of minerals of aluminum phosphate that caught and immobilized them for thousands of millions of years. This extracted information from the nature contributes guides and it allows 'to verify' the validity of the current proposals on the immobilization of radioactive wastes of a nuclear reactor. This work presents in clear and accessible form to the public in general on the secure 'design', operation, 'decommissioning' and 'storage' of the radioactive

  7. Oklo. A review and critical evaluation of literature

    International Nuclear Information System (INIS)

    Zetterstroem, Lena

    2000-10-01

    The Oklo natural fossil fission reactors in Gabon, Equatorial Africa, have been studied as a natural analogue for spent nuclear fuel in a geological environment. For these studies, it is important to know what has happened to these reactors since they formed. This review is focussed on existing geological and geochronological information concerning the Oklo reactors and the surrounding ore. A sequence of geological and geochemical events in the Oklo area, as described in the literature, is given. The data and the studies behind this established geochronology are discussed and evaluated. Of the regional geology, special attention is given to the dating of the Francevillian sediments, and the intrusion of a dolerite dyke swarm. The processes that led to the mineralisation at Oklo, the subsequent formation of the nuclear reactors and later migration of fission products are described. Further discussion concerns the studies of the dolerite dyke swarm, since this appears to be one of the most important events related to fission product migration. A close look at the data related to this event shows that further study of the age of the dolerite dykes, and their effect on the uraninite in the Oklo reactors, is needed

  8. Oklo working group meeting

    International Nuclear Information System (INIS)

    Von Maravic, H.

    1993-01-01

    Natural analogue studies have been carried out for several years in the framework of the European Community's R and D programme on radioactive waste; and within its recent fourth five-year programme on 'Management and storage of radioactive waste (1990-94)' the Community is participating in the Oklo study, natural analogue for transfer processes in a geological repository. The Oklo project is coordinated by CEA-IPSN (F) and involves laboratories from several CEA directorates (IPSN, DTA and DCC) which collaborate with other institutions from France: CREGU, Nancy; CNRS, Strasbourg and ENSMD, Fontainebleau. Moreover, institutes from non-EC member States are also taking part in the Oklo study. The second joint CEC-CEA progress meeting of the Oklo Working Group was held in April 1992 in Brussels and gave the possibility of reviewing and discussing progress made since its first meeting in February 1991 at CEA in Fontenay-aux-Roses. About 40 participants from 15 laboratories and organizations coming from France, Canada, Gabon, Japan, Sweden and the USA underline the great interest in the ongoing research activities. The meeting focused on the different tasks within the CEC-CEA Oklo project concerning (i) field survey and sampling, (ii) characterization of the source term, (iii) studies of the petrographical and geochemical system, and (iv) studies of the hydrogeological system and hydrodynamic modelling. (author) 17 papers are presented

  9. Oklo natural reactor. Study of uranium and rare earths migration on a core drilled through a reaction zone. Application to determination of the date of the nuclear reaction by measurement of fission products

    International Nuclear Information System (INIS)

    Ruffenach, J.C.

    1977-01-01

    Isotopic and chemical analysis of uranium and five rare earths: neodymium, samarium, europium, gadolinium and dysprosium were effected on fourteen samples taken in the same core drilled through a reaction zone of the Oklo uranium deposit. This study points out the general stability of uranium and fission rare earths; spatial distributions of these elements are quite analogous. Migrations have affected about 5% only of fission neodymium in the core of the reaction zone; corresponding values for samarium and gadolinium are slightly higher. These migration phenomena have carried rare earths to no more than 80 cm out of the core. By study of the europium it is shown that nuclear reactions have stayed in the ground since the time of reactions. On the other hand it is shown by analysis of the dysprosium that rare earths have not undergone an important movement. This study allow also the datation of nuclear reactions from the measurement of the quantity of fission neodymium produced. A value of 1.98x10 9 years is obtained slightly higher than the value obtained by geochronology [fr

  10. Natural fission reactors from Gabon. Contribution to the study of the conditions of stability of a natural radioactive wastes storage site (2 Ga)

    International Nuclear Information System (INIS)

    Pourcelot, L.

    1997-01-01

    The natural fission reactors of Oklo consists of a core of uraninite (60%) with fission products, embedded in a pure clay matrix. Thus, the aim of geological, mineral, and geochemical studies of the Oklo Reactors is to assess the behaviour of fission products in an artificial waste depository. Previous studies have shown that Reactor Zone 10, located in the Oklo mine, represents an example for an exceptional confinement of fission products since 2 Ga. In reactor Zone 9, located in Oklo open pit, migrations are more important. Reactor ZOne 13 was influenced by a thermal event due to a doleritic intrusion, located some twenty meters far away, one Ga years after fission reaction operations. In this study,we characterized temperature and redox conditions of fluids by using stable isotopes of uraninites and clays. Moreover mineralogical and chemical characteristics were defined. (author)

  11. New data on the geological environment of the natural reactors

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Besnus, Y.; Weber, F.

    1978-01-01

    Since the Libreville symposium in 1975 knowledge of the geological environment of the reactors has advanced as a result of a more extensive study of the Francevillian uranium deposits. In the Oklo deposit a detailed stratigraphy of the Cl bed (uraniferous mineralized bed) has been established, making it possible to re-establish stratigraphically the position of the natural reactors. A tectonic analysis of the Oklo deposit has revealed the special features of the Oklo structure and the reaction zones situated in the shear troughs. Petrographic studies have revealed the presence of two types of ore with distinct modes of formation. In the first case, the role played by organic materials seems predominant, while in the second case migrations of oxidizing solutions are the main source of the reconcentrations. Finally, a geochemical study made of samples from Oklo and Okelobondo points to the existence of an ''isolated'' geochemical phase containing uranium and a certain number of trace elements. This phase is associated with the organic material. This study also deals with the migration of lead at Oklo and Mounana. (author)

  12. The Oklo phenomenon as an analogue of radioactive waste disposal. A review

    International Nuclear Information System (INIS)

    Berzero, A.; D'Alessandro, M.

    1990-01-01

    This work demonstrates the utility of the Oklo uranium ore deposit and natural fission reactors as a long time scale analogue for man-made radioactive waste repositories. Oklo has opened a new horizon representing an unrivalled opportunity to apply isotopic geochemistry to the study of migrations of fission products after an extremely long cooling and storage time and to define the processes involved in the transport of these elements through geological materials. This is the topic of the first section of this report. In the second section the information available on retention or migration at Oklo of the most interesting fission products is presented trying to illustrate how relevant the Oklo experience is in formulating predictions on the destiny of high activity waste disposed of in stable geological formations

  13. Criticality in a high level waste repository. A review of some important factors and an assessment of the lessons that can be learned from the Oklo reactors

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1996-06-01

    The conditions and scenarios that might allow sufficient 239 Pu and/or 235 U to accumulate together with enough water to allow for moderation of neutron energies and thereby achieving a state where neutron-induced fission reactions could be sustained at a rate significantly above the natural rate of spontaneous fission is discussed. The uranium deposit in Oklo, Gabon, which was the site of naturally-occurring neutron-induced fission reactions approximately 2000 My ago is described. The chemistry, mineralogy, and conditions of the nuclear reactor operations are reviewed. Results of modelling the conditions for criticality at Oklo are used to estimate the amounts of spent fuel uranium that must be assembled in a favorable geometry in order to produce a similar reactive situation in a geologic repository. The amounts of uranium that must be transported and redeposited to reach a critical configuration are extremely large in relation to those that could be transported under any reasonably achievable conditions. In addition, transport and redeposition scenarios often require opposite chemical characteristics. It is concluded that the likelihood of achieving a critical condition due to accumulation of a critical mass of uranium outside the canisters after disposal is negligible. Criticality inside the canister is rendered impossible by the use of low-solubility materials inside the canisters that fill space and prevent the entry of enough water to allow moderation of neutron energies. Criticality due to plutonium outside the canister can be ruled out because it requires a series of processes, each of which has a vanishingly small probability. 25 refs, 9 tabs, 8 figs

  14. Tectonic analysis of the Oklo deposit

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Ruhland, M.; Weber, F.

    1975-01-01

    A large folded structure with a 40 0 incline and extending 500 m in the north-south direction has been uncovered at the Oklo mine. This structure has been analysed from the geometric and dynamic points of view in order to determine the possible role of tectonic activity in the creation of the uraniferous concentrations. Compression and extension zones which at certain points control the shape and arrangement of the lodes are associated with the structure. The natural reactors are situated in an extension zone where compartmentation and slippage, which explain the arrangement of the reactors, are observed

  15. Contribution of lead and thorium to the history of the Oklo reactors

    International Nuclear Information System (INIS)

    Devillers, C.; Menes, J.

    1978-01-01

    The authors report the results of measuring lead and thorium in a series of representative samples of the superconcentrations of uranium found in the Oklo mineralization. Interpretation of the data reveals the complexity of the history of lead in the deposit, but makes it possible to derive a number of important facts, namely early disturbance and recent, massive remobilization of lead. One is led to conclude that the date of the uranium emplacement may be greater than 1900 million years. The absence of high thorium contents in the ''normal'' rich ore confirms the importance of dating the nuclear reaction on the basis of the Th/U balance. This determination, which draws on the same set of data as for the Nd/U balance, gives a mean value close to 1930 million years. (author)

  16. Search for an ''Oklo Phenomenon'' in the Northeastern regions of Brazil

    International Nuclear Information System (INIS)

    Lima, F.W.; Vasconcellos, M.B.A.; Armelin, M.J.A.; Fulfaro, R.; Fulfaro, V.J.; Neves, B.B.B.

    1982-01-01

    Rocks samples from the Northeastern region of Brazil were analysed for their 235 U isotopic abundance, in search for the occurrence of an ''Oklo Phenomenon'' here. The samples were collected in locations that could have been connected to the African continent, according to the continental-drift theory, in accordance to the Francevillian formations in the Gabon Republic, in which place the Oklo natural fossil reactor is situated. Two methods were used for the determination of the 235 U abundance: activation analysis followed by high resolution gamma-ray spectrometry and activation analysis by delayed neutron counting. No evidence of 235 U depletion was found in the rock samples analysed. (author)

  17. Elementary migration around the Oklo nuclear reactors. Implications for high level radioactive wastes storage

    International Nuclear Information System (INIS)

    Menet-Dressayre, C.; Menager, M.T.

    1993-01-01

    The study of Uranium and rare earths near the reactors has displayed the radioelements transfer in the reactors neighbourhood. The main implications for high level radioactive wastes disposal in geological formations are discussed. 12 refs

  18. U.S. studies of the Oklo phenomenon

    International Nuclear Information System (INIS)

    Cowan, G.A.; Bryant, E.A.; Daniels, W.R.; Maeck, W.J.

    1975-01-01

    Analyses of samples from the Oklo site are in agreement with previous reports by Naudet et al concerning the occurrence of a natural chain reaction in Precambrian times. A preliminary modeling of the reactor suggests that the event took place 1.7 to 1.9 b.y. ago, and that the period of criticality lasted at least 2 x 10 5 years. Uranium may have accreted and, to a lesser extent, migrated during the period of criticality or subsequent to it. (U.S.)

  19. Investigations of the natural fission reactor program. Progress report, October 1977--September 1978

    International Nuclear Information System (INIS)

    Cowan, G.A.; Norris, A.E.

    1978-10-01

    The U.S. study of the Oklo natural reactor began in 1973 with the principal objectives of understanding the processes that produced the reactor and that led to the retention of many of its products. Major facets of the program have been the chemical separation and mass spectrometric analysis of the reactor components and products, the petrological and mineralogical examination of samples taken from the reactor zones, and an interdisciplinary modeling of possible processes consistent with reactor physics, geophysics, and geochemistry. Most of the past work has been on samples taken within the reactor zones. Presently, these studies give greater emphasis to the measurement of mobile products in additional suites of samples collected peripherally and ''downstream'' from the reactor zones. This report summarizes the current status of research and the views of U.S. investigators, with particular reference to the extensive work of the French scientists, concerning the main features of the Oklo natural fission reactor. Also mentioned briefly is the U.S. search for natural fission reactors at other locations

  20. Limits on cosmological variation of strong interaction and quark masses from big bang nucleosynthesis, cosmic, laboratory and Oklo data

    International Nuclear Information System (INIS)

    Flambaum, V.V.; Shuryak, E.V.

    2002-01-01

    Recent data on the cosmological variation of the electromagnetic fine structure constant from distant quasar (QSO) absorption spectra have inspired a more general discussion of the possible variation of other constants. We discuss the variation of strong scale and quark masses. We derive limits on their relative change from (i) primordial big bang nucleosynthesis, (ii) the Oklo natural nuclear reactor, (iii) quasar absorption spectra, and (iv) laboratory measurements of hyperfine intervals

  1. Parametric study of the criticality of natural reactors

    International Nuclear Information System (INIS)

    Naudet, R.

    1978-01-01

    Conditions for the criticality of natural reactors are investigated from a general point of view; a parametric study is presented, which expresses the possibility of chain reactions as functions of five parameters: the age of the deposit, the ore's uranium content, the volume of high-grade ore, the neutron capture of the vein of ore and the amount of water associated with the uranium. It is demonstrated that although criticality could theoretically be attained for ages that are not in excess of 1000 to 1200 MA, conditions would have to be exceptionally favorable for it since the deposits are clearly much younger than those at Oklo. The study offers a much better appreciation of the probability for discovery of other natural fissionable reactors

  2. Problems posed by the development of the Oklo phenomenon: tentative global interpretation

    International Nuclear Information System (INIS)

    Naudet, R.

    This paper discusses the basic problems posed by the development of the Oklo phenomenon: the conditions in which the reactions are triggered and propagated and how they have been controlled. The reactions were maintained by the destruction of neutron poisons in the ore and were controlled by temperature. Oklo is made up of a large number of contiguous reactors. Geological problems of the origin of the clays, desilification, and uranium concentration are discussed. Oklo is shown to be a very complex phenomenon which developed in space and time. Besides the thermal, neutron, and geochemical coupling, there is also a tectonic coupling

  3. Two billion year old natural analogs for nuclear waste disposal: the natural nuclear fission reactors in Gabon (Africa)

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.

    2002-01-01

    Two billion years ago, the increase of oxygen in atmosphere and the high 235 U/ 238 U uranium ratio (> 3%) made possible the occurrence of natural nuclear reactors on Earth. These reactors are considered to be a good natural analogue for nuclear waste disposal. Their preservation during such a long period of time is mainly due to the geological stability of the site, the occurrence of clays surrounding the reactors and acting as an impermeable shield, and the occurrence of organic matter that maintained the environment in reducing conditions, favourable for the stability of uraninite. Hydrogeochemical studies and modelling have shown the complexity of the geochemical system at Oklo and Bangombe (Gabon) and the lack of precise data about uranium and fission products retention and migration mechanisms in geological environments. (author)

  4. Okulo natural reactors

    International Nuclear Information System (INIS)

    Yamakawa, Minoru

    1993-01-01

    French CEA has reported in 1972 that natural nuclear reactors existed in Okulo uranium deposit in Gabon in Africa, that caused nuclear fission chain reaction (Okulo phenomena) spontaneously two billion years ago. The fission products and transuranic elements produced by the natural reactors have been preserved in strata without movement while subjected to geological phenomena for such very long years. 16 zones of the natural reactors have been discovered so far. The geological features of the Okulo uranium deposit are explained. The total amount of 235 U lost by the chain reaction was estimated to be about 6t, and the fission products were about 6t. The Okulo phenomena offered the valuable results of the synthetic formation disposal test that the nature has carried out for such long years. The significance of the study on natural analog is discussed. Organic substances and the mechanism of holding and movement of uranium and fission nuclides, the stability of uraninite and the age measurement of the deposit by Nd-Sm process are reported as the main results. (K.I.)

  5. Natural convection type reactor

    International Nuclear Information System (INIS)

    Nakayama, Takafumi; Horiuchi, Tetsuo; Moriya, Kimiaki; Matsumoto, Masayoshi; Akita, Minoru.

    1988-01-01

    Purpose: To improve the reliability by decreasing the number of dynamic equipments and safely shutdown the reactor core upon occurrence of accidents. Constitution: A pressure relief valve and a pressurizing tank or gravitational water falling tank disposed to the main steam pipe of a reactor are installed in combination. Upon loss-of-coolant accident, the pressure relief valve is opened to reduce the pressure in the reactor pressure vessel to the operation pressure for each of the tanks, thereby enabling to inject water in the pressurizing tank at first and, thereafter, water in the gravitational water falling tank successively to the inside of the pressure vessel. By utilizing the natural force in this way, the reliability can be improved as compared with the case of pumped water injection. Further, by injecting an aqueous boric acid to a portion of a plurality of tanks, if the control rod insertion becomes impossible, aqueous boric acid can be injected. (Takahashi, M.)

  6. Natural fission reactors in the Franceville basin, Gabon: A review of the conditions and results of a open-quotes critical eventclose quotes in a geologic system

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Holliger, P.; Blanc, P.L.

    1996-01-01

    Natural nuclear fission reactors are only known in two uranium deposits in the world, the Oklo and Bangombe deposits of the Franceville basin: Gabon. Since 1982, five new reactor zones have been discovered in these deposits and studied since 1989 in a cooperative European program. New geological, mineralogical, and geochemical studies have been carried out in order to understand the behavior of the actinides and fission products which have been stored in a geological environment for more than 2.0 Ga years. The Franceville basin and the uranium deposits remained geologically stable over a long period of time. Therefore, the sites of Oklo and Bangombe are well preserved. For the reactors, two main periods of actinide and radionuclides migration have been observed: during the criticality, under P-T conditions of 300 bars and 400-500 degrees C, respectively, and during a distention event which affected the Franceville basin 800 to 900 Ma ago and which was responsible for the intrusion of dolerite dikes close to the reactors. New isotopic analyses on uranium dioxides, clays, and phosphates allow us to determine their respective importance for the retention of fission products. The UO 2 matrix appears to be efficient at retaining most actinides and fission products such as REEs, Y, and Zr but not the volatile fission products (Cd, Cs, Xe, and Kr) nor Rb, Sr, and Ba. Some fissiogenic elements such as Mo, Tc, Ru, Rh, Pd, and Te could have formed metallic and oxide inclusion in the UO 2 matrix which are similar to those observed in artificial spent fuel. Clays and phosphate minerals also appear to have played a role in the retention of fissiogenic REEs and also of Pu. 82 refs., 21 figs., 12 tabs

  7. The nature of reactor accidents

    International Nuclear Information System (INIS)

    Domaratzki, Z.; Campbell, F.R.; Atchison, R.J.

    1981-01-01

    Reactor accidents are events which result in the release of radioactive material from a nuclear power plant due to the failure of one or more critical components of that plant. The failures, depending on their number and type, can result in releases whose consequences range from negligible to catastrophic. By way of examples, this paper describes four specific accidents which cover this range of consequence: failure of a reactor control system, loss of coolant, loss of coolant with impaired containment, and reactor core meltdown. For each a possible sequence of events and an estimate of the expected frequency are presented

  8. Search for other natural fission reactors

    International Nuclear Information System (INIS)

    Apt, K.E.; Balagna, J.P.; Bryant, E.A.; Cowan, G.A.; Daniels, W.R.; Vidale, R.J.

    1977-01-01

    Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m 3 and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater than 0.6 b.a. old and if they contained sufficient water but lacked neutron-absorbing impurities. While uranium deposits of northern Canada and northern Australia have received most attention, ore samples have been obtained from the following worldwide locations: the Shinkolobwe and Katanga regions of Zaire; Southwest Africa; Rio Grande do Norte, Brazil; the Jabiluka, Nabarlek, Koongarra, Ranger, and El Sharana ore bodies of the Northern Territory, Australia; the Beaverlodge, Maurice Bay, Key Lake, Cluff Lake, and Rabbit Lake ore bodies and the Great Bear Lake region, Canada. The ore samples were tested for isotopic variations in uranium, neodymium, samarium, and ruthenium which would indicate natural fission. Isotopic anomalies were not detected. Criticality was not achieved in these deposits because they did not have sufficient 235 U content (a function of age and total uranium content) and/or because they had significant impurities and insufficient moderation. A uranium mill monitoring technique has been considered where the ''yellowcake'' output from appropriate mills would be monitored for isotopic alterations indicative of the exhumation and processing of a natural reactor

  9. Developments in natural uranium - graphite reactors

    International Nuclear Information System (INIS)

    Bourgeois, J.

    1964-01-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  10. Synthesis of the mineralogic and petrographic studies of the ore minerals from Oklo, their gangues and the surrounding rocks

    International Nuclear Information System (INIS)

    Weber, F.; Geoffroy, J.; Le Mercier, M.

    Observations on the spot and mineralogical studies (reflection and transmission microscopy, x-ray examination, thermal analyses) have shown that the ore in the reaction zones differs from ''normal'' Oklo ore as regards both the nature of the mineralization and the gangue and the country rock. The relationship between the two ore types on one and between them and the country rock on the other is studied. Theories concerning the creation of the uraniferous deposit and the effects of subsequent changes due to diagenesis and recent weathering are discussed

  11. Summary of the mineralogical and petrographic studies of the Oklo ores, their gangues and the country rock

    International Nuclear Information System (INIS)

    Weber, F.; Geffroy, J.; Le Mercier, M.

    1975-01-01

    Observations on the spot and mineralogical studies (reflection and transmission microscopy, X-ray examination, thermal analyses) have shown that the ore in the reaction zones differs from ''normal'' Oklo ore as regards both the nature of the mineralization and the gangue and the country rock. The relationship between the two ore types on one hand and between them and the country rock on the other is studied. Theories concerning the creation of the uraniferous deposit and the effects of subsequent changes due to diagenesis and recent weathering are discussed

  12. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  13. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  14. Power limit and quality limit of natural circulation reactor

    International Nuclear Information System (INIS)

    Zhao Guochang; Ma Changwen

    1997-01-01

    The circulation characteristics of natural circulation reactor in boiling regime are researched. It is found that, the circulation mass flow rate and the power have a peak value at a mass quality respectively. Therefore, the natural circulation reactor has a power limit under certain technological condition. It can not be increased steadily by continually increasing the mass quality. Corresponding to this, the mass quality of natural circulation reactor has a reasonable limit. The relations between the maximum power and the reactor parameters, such as the resistance coefficient, the working pressure and so on, are analyzed. It is pointed out that the power limit of natural circulation reactor is about 1000 MW at present technological condition. Taking the above result and low quality stability experimental result into account, the authors recommend that the reasonable mass quality of natural circulation reactor working in boiling regime is from 2% to 3% under the researched working pressure

  15. PERFORMANCE IMPROVEMENT OF A CHEMICAL REACTOR BY NONLINEAR NATURAL OSCILLATIONS

    NARCIS (Netherlands)

    RAY, AK

    1995-01-01

    The dynamic behaviour of two coupled continuous stirred tank reactors in sequence is studied when the first reactor is being operated under limit cycle regimes producing self-sustained natural oscillations. The periodic output from the first reactor is then used as a forced input into the second

  16. Heat removal by natural convection in a RPR reactor

    International Nuclear Information System (INIS)

    Sampaio, P.A.B. de

    1987-01-01

    In this paper natural convection in RPR reactor is analysed. The effect of natural convection valves size on cladding temperature is studied. The reactor channel heat transfer problem is solved using finite elements in a two-dimensional analysis. Results show that two valves with Φ = 0.16 m are suited to keep coolant and cladding temperatures below 73 0 C. (author) [pt

  17. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  18. Natural occurring radioactive substances. Vol. 1

    Energy Technology Data Exchange (ETDEWEB)

    Emara, A E [National Center for radiation Research and Technology Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    Naturally occurring radioactive substances produced by cosmic rays of those of terrestrial origin are surveyed. The different radioactive decay series are discussed. Special emphasis is given to the element radium as regards its properties and distribution in different environmental samples. The properties of naturally occurring k-40 and its distribution in different natural media are also outlined. Induced radionuclides which are formed as a result of the interaction of cosmic rays with the constituents of the atmosphere are mentioned. In this respect the intensity of natural background radiation and the dose at different locations and levels is surveyed. Some regions of exceptionally high radioactivity which result in high exposure rates are mentioned. Monazite deposits and water springs are mentioned in some detail. The Oklo phenomenon as a natural reactor is also discussed. 8 tabs.

  19. Natural occurring radioactive substances. Vol. 1

    International Nuclear Information System (INIS)

    Emara, A.E.

    1996-01-01

    Naturally occurring radioactive substances produced by cosmic rays of those of terrestrial origin are surveyed. The different radioactive decay series are discussed. Special emphasis is given to the element radium as regards its properties and distribution in different environmental samples. The properties of naturally occurring k-40 and its distribution in different natural media are also outlined. Induced radionuclides which are formed as a result of the interaction of cosmic rays with the constituents of the atmosphere are mentioned. In this respect the intensity of natural background radiation and the dose at different locations and levels is surveyed. Some regions of exceptionally high radioactivity which result in high exposure rates are mentioned. Monazite deposits and water springs are mentioned in some detail. The Oklo phenomenon as a natural reactor is also discussed. 8 tabs

  20. Startup method for natural convection type nuclear reactor

    International Nuclear Information System (INIS)

    Utsuno, Hideaki.

    1993-01-01

    In a nuclear reactor started by natural convection, no sufficient stability margin can be ensured upon start up. Then, in the present invention, a deaerating operation is conducted before start-up of the reactor, then control rods are withdrawn after conducting the deaerating operation and temperature and pressure are raised by nuclear heating, to obtain a rated power. As a result, reactor power and subcooling at the inlet of the reactor core are within a range of lower than a geysering forming region, thereby enabling to prevent occurence of geysering inherent to the start-up of operation in a natural convection state, shorten the start-up time, as well as remove oxygen dissolved in coolants. (N.H.)

  1. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  2. Development of thermohydraulic software for PWR reactors with natural circulation

    International Nuclear Information System (INIS)

    Chasseur, Alfredo F.; Rauschert, A.; Delmastro, Dario F.

    2009-01-01

    The basics concepts about the development of software for steady state analysis of a reactor with natural circulations, in the primary circuit, are exposed. The reactor type is pressurized light water. The equations, correlations and flux diagrams of the source code of the software developed are shown. The source code of the software was written in FORTRAN 77 making use of modular technique, this save development effort and release of news versions is simplified. (author)

  3. International benchmark on the natural convection test in Phenix reactor

    International Nuclear Information System (INIS)

    Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

    2013-01-01

    Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

  4. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  5. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  6. Natural circulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Bastos, J.L.F.; Loureiro, L.V.; Rocha, R.T.V. da; Umbehaun, P.E.

    1992-01-01

    Several analytical modelling have been done for steady-state and slow transients conditions, besides more sophisticated studies considering two and three dimensional effects in a very simple geometry. Under severe accident conditions for PWR a code to analyse natural circulation has been developed by Westinghouse. This paper discusses the problem of natural circulation in a complex geometry similar to that of nuclear power plants. A first experiment has been done at the integral test facility of 'Co-ordination of Special Projects-Ministry of Naval Affairs' (Coordenadoria para Projetos Especiais -Ministerio da Marinha, COPESP) for several flux conditions. The results obtained were compared with numerical simulations for the steady-state regime. 09 refs, 05 figs, 01 tab. (B.C.A.)

  7. The light water natural uranium reactor

    International Nuclear Information System (INIS)

    Radkowsky, A.

    A new type of light water seed blanket with the seed having 20% enrichment and the blanket a special combination of elements of natural uranium and thorium, relatively close packed, but sufficient spacing for heat transfer purpose is described. The blanket would deliver approximately half the total energy for about 10,000 MWDIT, so this type of core would be just as economical or better in uranium ore consumation as present cores. (author)

  8. The Oklo phenomenon and the role of nuclear data in its study

    International Nuclear Information System (INIS)

    Roth, E.; Hagemann, R.; Ruffenach.

    1979-01-01

    The study of the OKLO phenomenon requires, but also provides, in some cases better nuclear data than that were available at the time of its discovery. Examples of this situation are given particularly for rare earth's fission yields and neutron capture cross sections

  9. Study on operational aspect of natural circulation HLMC reactor (1)

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

    2000-08-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

  10. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  11. Nuclear reactor lid cooling which can work by natural circulation

    International Nuclear Information System (INIS)

    Wagner, J.

    1985-01-01

    The well-known air cooling of the lid of liquid metal cooled nuclear reactors is improved by the start of natural convection flow ensuring removal of heat in a sufficiently short time, if the blower fails. Go and return branches of the individual cooling circuits are arranged at different heights for this purpose. The circulation is supported by opening valves, which provide a direct path into the reactor building for the cooling air. The draught can be increased by setting up special chimneys. The start of circulation is aided by the temporary opening of another valve. (orig.) [de

  12. Natural gas turbine topping for the iris reactor

    International Nuclear Information System (INIS)

    Oriani, L.; Lombardi, C.; Paramonov, D.

    2001-01-01

    Nuclear power plant designs are typically characterized by high capital and low fuel costs, while the opposite is true for fossil power generation including the natural gas-fired gas turbine combined cycle currently favored by many utilities worldwide. This paper examines potential advantages of combining nuclear and fossil (natural gas) generation options in a single plant. Technical and economic feasibility and attractiveness of a gas turbine - nuclear reactor combined cycle where gas turbine exhaust is used to superheat saturated steam produced by a low power light water reactor are examined. It is shown that in a certain range of fuel and capital costs of nuclear and fossil options, the proposed cycle offers an immediate economic advantage over stand-alone plants resulting from higher efficiency of the nuclear plant. Additionally, the gas turbine topping will result in higher fuel flexibility without the economic penalty typically associated with nuclear power. (author)

  13. Natural gas turbine topping for the iris reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oriani, L.; Lombardi, C. [Politecnico di Milano, Milan (Italy); Paramonov, D. [Westinghouse Electric Corp., LLC, Pittsburgh, PA (United States)

    2001-07-01

    Nuclear power plant designs are typically characterized by high capital and low fuel costs, while the opposite is true for fossil power generation including the natural gas-fired gas turbine combined cycle currently favored by many utilities worldwide. This paper examines potential advantages of combining nuclear and fossil (natural gas) generation options in a single plant. Technical and economic feasibility and attractiveness of a gas turbine - nuclear reactor combined cycle where gas turbine exhaust is used to superheat saturated steam produced by a low power light water reactor are examined. It is shown that in a certain range of fuel and capital costs of nuclear and fossil options, the proposed cycle offers an immediate economic advantage over stand-alone plants resulting from higher efficiency of the nuclear plant. Additionally, the gas turbine topping will result in higher fuel flexibility without the economic penalty typically associated with nuclear power. (author)

  14. Natural scientific strategy for nuclear wastes: Philosophy of Dr. Rodney C. Ewing at Stanford

    Energy Technology Data Exchange (ETDEWEB)

    Woo, T. H. [The Cyber University of Korea, Seoul (Korea, Republic of)

    2016-10-15

    In the geological study, the naturally adapted region was discovered in Oklo, a central African area of Gabon seen in Fig. 1. Currently, this naturally operated nuclear fission site has been stabilized where the artificial shielding structure was not equipped. So, it has been studied to be an exemplified high-level nuclear waste repository. This could give a lesson that one should make the nuclear repository construction. The simplified configuration of the Oklo natural nuclear reactor site is described. The important mentions of Dr. Ewing reflected his philosophy in geological repository and there is a biography. It is necessary to understand the very big difference between human’s cognitive timescale and geological timescale where it could be nearly impossible to construct the long-term or even permanent nuclear waste repository. It is important to understand the opposite opinions of the nuclear waste repository. The government made the milestone of the South Korean repository policy in which the final repository completion was decided as 2053, about 40 years later. Until the time, there is enough time to rethink for the long-term waste repository. It is certain the policy-maker of this nuclear waste policy will be out of the nuclear community before 2053. This means the politician will not take any responsibility of the result of this policy.

  15. Natural scientific strategy for nuclear wastes: Philosophy of Dr. Rodney C. Ewing at Stanford

    International Nuclear Information System (INIS)

    Woo, T. H.

    2016-01-01

    In the geological study, the naturally adapted region was discovered in Oklo, a central African area of Gabon seen in Fig. 1. Currently, this naturally operated nuclear fission site has been stabilized where the artificial shielding structure was not equipped. So, it has been studied to be an exemplified high-level nuclear waste repository. This could give a lesson that one should make the nuclear repository construction. The simplified configuration of the Oklo natural nuclear reactor site is described. The important mentions of Dr. Ewing reflected his philosophy in geological repository and there is a biography. It is necessary to understand the very big difference between human’s cognitive timescale and geological timescale where it could be nearly impossible to construct the long-term or even permanent nuclear waste repository. It is important to understand the opposite opinions of the nuclear waste repository. The government made the milestone of the South Korean repository policy in which the final repository completion was decided as 2053, about 40 years later. Until the time, there is enough time to rethink for the long-term waste repository. It is certain the policy-maker of this nuclear waste policy will be out of the nuclear community before 2053. This means the politician will not take any responsibility of the result of this policy

  16. Automated scoping methodology for liquid metal natural circulation small reactor

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2014-01-01

    Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

  17. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  18. Industrial and natural nuclear reactors; Industrielle und natuerliche Kernreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Binnewies, Michael [Hannover Univ. (Germany); Willner, Helge; Woenckhaus, Juergen

    2015-08-15

    As described in the preceding article, all elements with atomic masses above that of iron and also the radioactive elements thorium and uranium have been formed by a supernova star explosion. Their long-lived isotopes of thorium and uranium are now distributed in the earth crust. The chemistry of uranium and thorium is of less importance, but these elements can be used to produce enormous amounts of energy in nuclear power stations. It will be described how it works. Surprisingly, small natural nuclear reactors were producing heat during hundreds of thousand years. Subsequently, we are dealing with this phenomenon, the principle of nuclear fission, the different types of nuclear reactors, security aspects and new developments.

  19. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  20. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au)

  1. Natural Circulation Capability Assessments for a Small-medium Reactor

    International Nuclear Information System (INIS)

    Choi, Sun Do

    2010-02-01

    Small-medium reactors have been highly evaluated to have more safe characteristics than those of large reactors. In addition, it could be used for a variety of purposes, such as small-scale power production in mountainous of island area, seawater desalination, regional heating system. For a higher safety, studies about a way of using natural circulation have being conducted around world. CAREM(Argentina), AST- 500(Russia), and NHR-200(china) etc. According to this tendency, REX- 10(Regional Energy rX-10) is designed in Korea for regional heating and small-scale power production. To investigate the thermal-hydraulic behavior of REX-10, we designed Rex-10 Test Facility (RTF), simulating REX-10, by using the scaling law. The scaling ratios of length, volume and power were set with 1/1, 1/50 and 1/50, respectively. The diameter and total length of RTF are 40 cm and approximately 6 m, respectively. The facility is composed of various components, which are a core in the bottom part, a heat exchanger in the middle part, a pressurizer and hot legs in the upper part, and chillers outside the facility. The test instrumentation is also designed to measure temperatures, flow rates, pressures, and pressure drop. The experiment parameters were adopted based on the 1-dimensional approach. There are a variety of parameters which influence natural circulation behavior such as heater power, overall flow resistance parameter, the distance between the center of the heat exchanger and the core. As the experimental geometries are fixed, it is found that the most important parameter is the heater power under the experimental conditions. In addition, to evaluate the effect of heater power, some experiments were conducted at varying heater power condition (from 70 kW to 170 kW) under constant primary pressure (2.0 MPa) and secondary flow rate (4.5 liter per minute). As the results of the experiments, the temperature and flow rate increase with increasing heater power. The flow rate is

  2. Evidence for a Large Natural Nuclear Reactor in Mars Past

    Science.gov (United States)

    Brandenburg, J. E.

    2006-05-01

    It has long been known that The isotopic ratios 129 Xe/132Xe and 40Ar/36Ar are very high in Mars atmosphere relative to Earth or meteoritic backgrounds. This fact has allowed the SNC meteorites to be identified as Martian based on their trapped gases (1). However, while the isotopic anomalies explained one mystery, the origin of the SNC meteorites, they created a new mystery: the rock samples from Mars show no evidence of the large amounts of Iodine or Potassium that would give naturally give rise to the Xenon and Argon isotopic anomalies (2). In fact, the Martian meteorites are depleted in Potassium relative to earth rocks. This is added to the fact that for other isotopic systems such as 80Kr, Mars rock samples must be irradiated by neutrons at fluences of 1015 /cm2 to explain observed abundances (1) . Compounding the mystery is the fact that Mars surface layer has elevated levels of Uranium and Thorium relative to Earth and even its own rocks, as determined from SNCs (3). These anomalies can be explained if some large nuclear energy release, such as by natural nuclear reactors known to have operated on Earth (4) in in some concentrated ore body, occurred with perhaps a large volcano like explosion that spread residues over the planets surface. Based on gamma ray observations from orbit (3), and the correlations of normally uncorrelated Th and K deposits , the approximate location of this event would appear to have been in the north of Mars in a region in Acidalia Planitia centered at 45N Latitude and 15W Longitude (5). The possibility of such a large radiological event in Mars past adds impetus to Mars exploration efforts and particularly to a human mission to Mars to learn more about this possible occurrence. (1) Swindle, T. D. , Caffee, M. W., and Hohenberg, C. M., (1986) "Xenon and other Noble Gases in Shergottites" Geochimica et Cosmochimica Acta, 50, pp 1001-1015. (2) Banin, A., Clark, B.C., and Wanke, H. "Surface Chemistry and Mineralogy" (1992) in "Mars

  3. Natural-draught cooling tower of the Philippsburg-1 reactor

    International Nuclear Information System (INIS)

    Ernst, G.; Wurz, D.

    1983-01-01

    In spring 1980 a comprehensive research programm was carried out on the natural-draught cooling tower of the Philippsburg-1 reactor. The study was meant to synchronously acquire all parameters necessary for the evaluation of plant operation and cooling tower emissions. The study is subdivided into 8 sub-projects. Parts 1 to 7 that are included in this progress-of-work report describe experimental work and discuss the results. A critical analysis of measuring results proves that the values for operational behaviour and cooling tower emissions were duly anticipated. Even a very critical judgment of the results can exclude direct or indirect hazards for humans, animals and plants owing to cooling tower emissions. Sub-project 8 compares results from diffusion calculations (24 models) to results gained from experiments. The results of sub-project 8 will be published in a progress report to come. (orig.) [de

  4. SOILS AS NATURAL REACTORS FOR SWINE WASTEWATER TREATMENT

    Directory of Open Access Journals (Sweden)

    Francisco Bautista

    2011-04-01

    Full Text Available The ability of soils to mineralize organic matter depends on their individual characteristics; when waste waters are added to them their organic matter content (OM, cationic exchange capacity (CEC and percentage of clay (PC are altered. Pedotransfer functions (PTF enable certain processes to be determined from easily measured soil properties. The aims of this study were i to generate PTF to estimate the retention and mineralisation of dissolved organic matter (DOM present in swine wastewater (SWW based on measurements of OM, CEC and PC and ii to identify the soils most suited to acting as natural reactors for treating SWW, using multicriteria analysis. Samples were taken from ten soils (epipedons or superficial samples to measure the retention of dissolved organic matter (RDOM in 30 cm high soil columns, making three applications of SWW. In addition, an experiment was carried out in pots to measure the effect of SWW on soil carbon evolution (SCE and the potential anaerobic nitrogen mineralisation (PANM. Multiple regressions were made using soil OM (%, CEC (cmol+ kg-1 and PC (% as independent variables and Chemical Oxygen Demand (COD, SCE and PANM as dependent variables. The PFT found were RDOM = 41.5 + (2.8*CEC – (0.81*PC – (3.5*OM  r= 0.81; SCE =  542.3 + (20.1*OM + (4.6*CEC – (2.7*PC r= 0.96; PANM = -8.4 + (3.45*OM + (1.12*PC – (2.20*CEC r= 0.88. The most suitable soils for acting as natural reactors of SWW were the Luvisol LVct and an unclassified EPI-1. Â

  5. Natural repository analogue program. Progress report, July 1-September 30, 1981

    International Nuclear Information System (INIS)

    Curtis, D.B.

    1982-03-01

    A report on the immobilization of uranium in the earth's crust has been completed. Techniques have been developed to do a comprehensive mass inventory of the Oklo reactor zones. These techniques were applied to a compilation of data from Oklo zones 2 and 3-4. The study shows large deficiencies of neodymium, ruthenium, and mass 99 elements ( 99 Tc or 99 Ru) in the reactor zones. The extent of these deficiencies are correlated with the intensity of the nuclear reactions. Analyses of ores from the Key Lake uranium mineralization show that 60 to 70% of the radiogenic lead is missing from the ores

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

    International Nuclear Information System (INIS)

    Pinedo V, J.L.

    1979-01-01

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

  7. FFTF primary system transition to natural circulation from low reactor power

    International Nuclear Information System (INIS)

    Bouchey, G.D.; Additon, S.L.; Nutt, W.T.

    1980-01-01

    Plans for reactor and primary loop natural circulation testing in the Fast Flux Test Facility (FFTF) are summarized. Detailed pretest planning with an emphasis on understanding the implications of process noise and model uncertainties for model verification and test acceptance are discussed for a transition to natural circulation in the reactor core and primary heat transport loops from initial conditions of 5% of rated reactor power and 75% of full flow

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

    International Nuclear Information System (INIS)

    Djati Hoesen Salimy

    2010-01-01

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

  9. Reactor modeling and process analysis for partial oxidation of natural gas

    NARCIS (Netherlands)

    Albrecht, B.A.

    2004-01-01

    This thesis analyses a novel process of partial oxidation of natural gas and develops a numerical tool for the partial oxidation reactor modeling. The proposed process generates syngas in an integrated plant of a partial oxidation reactor, a syngas turbine and an air separation unit. This is called

  10. BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G. [N.A. Dollezhal Institute ' NIKIET' , PO Box 788, Moscow, 101000 (Russian Federation)

    2006-07-01

    This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with {beta}{sub eff}, low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

  11. BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

    International Nuclear Information System (INIS)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G.

    2006-01-01

    This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with β eff , low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

  12. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  13. Radioactive Contamination Near Natural Uranium - Graphite - Gas Reactors

    International Nuclear Information System (INIS)

    Chassany, J.; Pouthier, J.

    1967-01-01

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

  14. Self-pressurization analysis of the natural circulation integral nuclear reactor using a new dynamic model

    Directory of Open Access Journals (Sweden)

    Ali Farsoon Pilehvar

    2018-06-01

    Full Text Available Self-pressurization analysis of the natural circulation integral nuclear reactor through a new dynamic model is studied. Unlike conventional pressurized water reactors, this reactor type controls the system pressure using saturated coolant water in the steam dome at the top of the pressure vessel. Self-pressurization model is developed based on conservation of mass, volume, and energy by predicting the condensation that occurs in the steam dome and the flashing inside the chimney using the partial differential equation. A simple but functional model is adopted for the steam generator. The obtained results indicate that the variable measurement is consistent with design data and that this new model is able to predict the dynamics of the reactor in different situations. It is revealed that flashing and condensation power are in direct relation with the stability of the system pressure, without which pressure convergence cannot be established. Keywords: Condensation Power, Flashing Phenomenon, Natural Circulation, Self-Pressurization, Small Modular Reactor

  15. A potential of boiling water power reactors with a natural circulation of a coolant

    International Nuclear Information System (INIS)

    Osmachkin, V.S.; Sokolov, I.N.

    1998-01-01

    The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

  16. Aspects on optimization of natural uranium fuel utilization in heavy water reactors

    International Nuclear Information System (INIS)

    1978-08-01

    This paper is dealing with a possibility to decrease the natural uranium consumption of CANDU PHWR using the once-through cycle. This possibility is based on the utilization of slightly enriched uranium. The optimal two-zone structure of a reactor using natural uranium is found out. The optimal criterium is the maximization of the burnup (equivalent to minimization of uranium requirements) with a constraint on power density radial uniformity factor. As regards the enriched uranium, the optimal enrichment and the two-zone structure of a reactor which minimizes the natural uranium requirement with constraints on uniformity factor and maximum burnup are established. Corresponding to a maximum burnup of 16,000 MWd/t and 1% enrichment, the natural uranium requirement is found to be 10% less than that of the natural uranium reactor

  17. Power optimization in the STAR-LM modular natural convection reactor system. Topic 2.1 advanced reactor power plants

    International Nuclear Information System (INIS)

    Spencer, B.W.; Sienicki, J.J.; Farmer, M.T.

    2001-01-01

    The secure, transportable, autonomous reactor (STAR) project addresses the needs of developing countries and independent power producers for a small (300 MWt), multi-purpose energy system. The STAR-LM variant described here is a liquid metal cooled, fast spectrum reactor system. Previous development of a reference STAR-LM design resulted in a 300 MWt modular, pool- type reactor based on criteria for factory fabrication of modules, full transportability of modules (barge, rail, overland), fast construction and startup, and semi-autonomous operation. Earlier work on the reference 300 MWt concept focused first on addressing whether 100% natural circulation heat transport was achievable under the module size constraints for full transportability and under the coolant and cladding peak temperature limitations imposed by the existing Russian database for ferritic-martensitic core material with oxide-layer corrosion protection. Secondly, owing to uncertainties and limitations in the available Russian materials compatibility database, the objective of the reference design was to address how low the coolant and cladding peak temperatures could be commensurate with achieving 300 MWt power level with 100% natural circulation in a fully transportable module size. In the present work we have refocused the approach to attempt to maximize the power achievable in the reactor module based on preserving the criteria for full module transportability and remaining within the materials compatibility database limits. (author)

  18. The relationship between natural uranium and advanced fuel cycles in CANDU reactors

    International Nuclear Information System (INIS)

    Lane, A.D.; McDonnell, F.N.; Griffiths, J.

    1988-11-01

    CANDU is the most uranium-economic type of thermal power reactor, and is the only type used in Canada. CANDU reactors consume approximately 15% of Canadian uranium production and support a fuel service industry valued at ∼$250 M/a. In addition to their once-through, natural-uranium fuel cycle, CANDU reactors are capable of operating with slightly-enriched uranium (SEU), uranium-plutonium and thorium cycles, more efficiently than other reactors. Only SEU is economically attractive in Canada now, but the other cycles are of interest to countries without indigenous fuel resources. A program is underway to establish the fuel technologies necessary for the use of SEU and the other fuel cycles in CANDU reactors. 22 refs

  19. Fast breeder reactors insertion in a D2O - natural U nuclear power plants park

    International Nuclear Information System (INIS)

    Gho, C.J.

    1985-01-01

    A model for the evolution of Argentine's installed nuclear power for the next 40 years is presented. The consequences of fast breeder reactors' introduction are studied in both autarchic Pu cycle and a limited reprocessing system. The passage of a reactor park like the national, of natural U - heavy water to one of fast breeder reactors, can only be obtained in a very long term due, fundamentally, to the need of Pu produced for those to feed the last ones. (M.E.L.) [es

  20. Natural circulation cooling in US pressurized water reactors

    International Nuclear Information System (INIS)

    Berta, V.T.; Wilson, G.E.; Boyack, B.E.

    1989-01-01

    The research into the modes of, and heat removed by, natural circulation in PWR systems is reviewed for the purpose of determining the status of this method for off-nominal recovery procedures. The referenced information comes from all facets of the nuclear industry, both domestic and international. The information focuses on recent research (1986--1988); however, pre-1986 research is summarized and referenced. Particular attention is paid to the role of scaling in the experimental facilities and analytical tools. Three modes of natural-circulation cooling are covered: condensation. The conclusion of the review is that the new research reconfirms the pre-1986 conclusion that natural circulation is a viable means of decay heat removal. In addition, the new research sufficiently completes the acquisition of an appropriate experimental data base and the development of system codes to permit the design of valid plant recovery procedures incorporating all three modes of natural circulation. 48 refs., 1 fig., 3 tabs

  1. Analysis of SBO accident and natural circulation of 49-2 swimming pool reactor

    International Nuclear Information System (INIS)

    Wu Yuanyuan; Liu Tiancai; Sun Wei

    2012-01-01

    The transient thermal hydraulic characteristics of 49-2 Swimming Pool Reactor (SPR) were analyzed by RELAP5/MOD3.3 code to verify the capability of natural circulation and minus reactivity feedback for accident mitigation under the condition of station blackout (SBO). Then, the effects on accident consequence and sequence for core channels and primary pumps were briefly discussed. The calculation results show that the reactor can be shutdown by the effect of minus reactivity feedback, and the residual heat can be removed through the stable natural circulation. Therefore, it demonstrates that the 49-2 SPR is safe during the accident of SBO. (authors)

  2. Survey of natural-circulation cooling in U.S. pressurized water reactors

    International Nuclear Information System (INIS)

    Boyack, B.E.

    1985-01-01

    Literature describing natural circulation analyses, experiments, and plant operation have been obtained from the Nuclear Regulatory Commission, reactor vendors, utility-sponsored research groups, utilities, national laboratories, and foreign sources. These have been reviewed and significant results and conclusions identified. Three modes of natural-circulation cooling are covered: single phase, two-phase, and reflux condensation. Single-phase natural circulation is amply verified by plant operational data, test data from scaled experimental facilities, and analysis with assessed computer codes. Ample evidence also exists that two-phase natural circulation can successfully cool pressurized water reactors. This mode occurs during certain events such as small-break loss-of-coolant accidents. The data base for reflux condensation is primarily from tests in scaled experimental facilities. There are no plant operational data and only limited assessment of thermal-hydraulic systems codes has been performed. Further work is needed before this mode of natural circulation can be confidently used

  3. A lumped parameter core dynamics model for MTR type research reactors under natural convection regime

    International Nuclear Information System (INIS)

    Ardaneh, Kazem; Zaferanlouei, Salman

    2013-01-01

    Highlights: ► A model is presented to simulate the reactivity insertion transient in MTR reactors. ► Transient dynamics of IAEA 10 MW MTR type research reactor are evaluated. ► Maximum unprotected reactivity insertion for safe condition is calculated. ► The model predictions are validated with corresponding results in the literature. - Abstract: On the basis of lumped parameter modeling of both the kinetic and thermal–hydraulic effects, a reasonably accurate simplified model has been developed to predict the dynamic response of MTR reactors following to an unprotected reactivity insertion under natural convection regime. By this model the reactor transient behavior at a given initial steady-state can be solved by a set of ordinary differential equations. The model predictions have an acceptable consent with corresponding results of reactivity insertion transients analyzed in the literature. The inherent safety characteristics of MTR research reactors utilizing natural convection is clearly demonstrated by the expanded model. The safety margin of reactor operating is selected ONB condition and thereby the proposed model determines that any slight increase in the value of $0.73 for inserted reactivity will cause the maximum cladding surface temperature to exceed the ONB condition

  4. Operational and passive safety aspects of the STAR-LM natural convection HLMC reactor. Study on operational aspects of a natural circulation HLMC reactor. 2

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Petkov, P.V.

    2001-09-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100+% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100+% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the earlier Phase 1 of the project, the stage for the overall study was prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code was developed that has the capabilities to calculate operational and accident transients. Code input was prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change-in-turbine load demonstrated the capability to analyze typical transient cases. The present second phase of the study involves documentation of the plant dynamics analysis computer code including major assumptions and thermal hydraulic equations as well as application of the code to calculate operational transients and postulated accidents. The following normal and accident scenarios are calculated: initial startup; normal shutdown; startup from hot standby; decrease-in-turbine load; increase-in-turbine load; loss-of-heat sink without scram; overcooling event without scram; and unprotected transient overpower. For the decrease

  5. Effect of ship motions and flow stability in a small marine reactor driven by natural circulation

    International Nuclear Information System (INIS)

    Yoritsune, Tsutomu; Ishida, Toshihisa

    2001-12-01

    By using a small reactor as a power source for investigations and developments under sea, widely expanded activity is expectable. In this case, as for a nuclear reactor, small-size and lightweightness, and simplification of a system are needed with the safety. In JAERI, very small reactors for submersible research vessel (Deep-sea Reactor DRX and submersible Compact Reactor SCR) have been designed on the basis of needs investigation of sea research. Although the reactor is a PWR type, self-pressurization and natural circulation system are adopted in a primary system for small size and lightweightness. The fluid flow condition of the reactor core outlet is designed to be the two-phase with a low quality. Although the flow of a primary system is the two-phase flow with a low quality, the density wave oscillation may occur according to operating conditions. Moreover, since there are ship motions of heaving (the vertical direction acceleration) etc., when a submersible research vessel navigates on the sea surface, the circulation flow of the primary system is directly influenced by this external force. In order to maintain stable operations of the reactor, it is necessary to clarify effects of the flow stability characteristic of the primary coolant system and the external force. Until now, as for the flow stability of a nuclear reactor itself, many research reports have been published including the nuclear-coupled thermal oscillation of BWRs such as LaSalle-2, WNP-2 etc. As for the effect of external force, it is reported that the acceleration change based on a seismic wave affects the reactor core flow and the reactor power in a BWR. On the other hand, also in a PWR, since adoption of natural circulation cooling is considered for a generation 4 reactor, it is thought that the margin of the reactor core flow stability becomes an important parameter in the design. The reactor coolant flow mentioned in this report is the two-phase natural circulation flow coupled with

  6. A simplified model of aerosol removal by natural processes in reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Powers, D.A.; Washington, K.E.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States); Burson, S.B. [Nuclear Regulatory Commission, Washington, DC (United States)

    1996-07-01

    Simplified formulae are developed for estimating the aerosol decontamination that can be achieved by natural processes in the containments of pressurized water reactors and in the drywells of boiling water reactors under severe accident conditions. These simplified formulae were derived by correlation of results of Monte Carlo uncertainty analyses of detailed models of aerosol behavior under accident conditions. Monte Carlo uncertainty analyses of decontamination by natural aerosol processes are reported for 1,000, 2,000, 3,000, and 4,000 MW(th) pressurized water reactors and for 1,500, 2,500, and 3,500 MW(th) boiling water reactors. Uncertainty distributions for the decontamination factors and decontamination coefficients as functions of time were developed in the Monte Carlo analyses by considering uncertainties in aerosol processes, material properties, reactor geometry and severe accident progression. Phenomenological uncertainties examined in this work included uncertainties in aerosol coagulation by gravitational collision, Brownian diffusion, turbulent diffusion and turbulent inertia. Uncertainties in aerosol deposition by gravitational settling, thermophoresis, diffusiophoresis, and turbulent diffusion were examined. Electrostatic charging of aerosol particles in severe accidents is discussed. Such charging could affect both the coagulation and deposition of aerosol particles. Electrostatic effects are not considered in most available models of aerosol behavior during severe accidents and cause uncertainties in predicted natural decontamination processes that could not be taken in to account in this work. Median (50%), 90 and 10% values of the uncertainty distributions for effective decontamination coefficients were correlated with time and reactor thermal power. These correlations constitute a simplified model that can be used to estimate the decontamination by natural aerosol processes at 3 levels of conservatism. Applications of the model are described.

  7. Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong Woon

    2009-01-01

    This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

  8. Natural convection as the way of heat removal from fast reactor core at cooldown regimes

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Kuzina, J.A.; Uhov, V.A.; Sorokin, G.A.

    2000-01-01

    The problems of thermohydraulics in fast reactors at cooldown regimes at heat removal by natural convection are considered The results of experiments and calculations obtained in various countries in this area are presented. The special attention is given to heat removal through inter-assembly space in the core and also to problems of thermohydraulics in the upper plenum. (author)

  9. Improved locations of reactivity devices in future CANDU reactors fuelled with natural uranium or enriched fuels

    International Nuclear Information System (INIS)

    Boczar, P.G.; Van Dyk, M.T.

    1987-02-01

    A new configuration of reactivity devices is proposed for future CANDU reactors which improves the core characteristics with enriched fuels, while still allowing the use of natural uranium fuel. Physics calculations for this new configuration are presented for four fuel types: natural uranium, mixed plutonium - uranium oxide (MOX) having a burnup of 21 MWd/kg, and slightly enriched uranium (SEU) having burnups of either 21 or 31 MWd/kg

  10. Bacteria, colloids and organic carbon in groundwater at the Bangombe site in the Oklo area

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, K. [ed.

    1996-02-01

    This report describes how microorganisms, colloids and organic matter were sampled from groundwater from six boreholes at the Bangombe site in the Oklo region and subsequently analyzed. For analysis of microorganisms, DNA was extracted from groundwater, amplified and cloned and information available in the ribosomal 16S rRNA gene was used for mapping diversity and distribution of bacteria. Each borehole was dominated by species that did not dominate in any of the other boreholes, a result that probably reflects documented differences in the geochemical environment. Analyses of sampled colloids included SEM and ICP-MS analysis of colloids on membrane and single particle analysis of samples in bottles. The colloid concentration was rather low in these Na-Mg-Ca-HCO{sub 3} type waters. Trace element results show that transition metals and some heavy metals are associated with the colloid phase. Distribution coefficients of trace elements between the water and colloid phases were estimated. For example for uranium, an average of 200 pg/ml was detected in the water, and 40 pg/ml was detected in the colloid phase. A K{sub p} value of 2* 10{sup 6} ml/g was calculated, considering (colloid) = 100 ng/ml. Groundwater samples were collected for analysis of the concentration of organic carbon (TOC), humic substances and metals associated with the humic substances. TOC varied in the range 4-14 mg/l in three boreholes, one borehole had a TOC<1.5 mg/l. The metal speciation study indicated that a large fraction, 8-67% of uranium was bound to the humic matter compared to the fractions of Ca and Fe (<0.4% and 0.02-10%, resp.). 60 refs, 8 figs, 16 tabs.

  11. Bacteria, colloids and organic carbon in groundwater at the Bangombe site in the Oklo area

    International Nuclear Information System (INIS)

    Pedersen, K.

    1996-02-01

    This report describes how microorganisms, colloids and organic matter were sampled from groundwater from six boreholes at the Bangombe site in the Oklo region and subsequently analyzed. For analysis of microorganisms, DNA was extracted from groundwater, amplified and cloned and information available in the ribosomal 16S rRNA gene was used for mapping diversity and distribution of bacteria. Each borehole was dominated by species that did not dominate in any of the other boreholes, a result that probably reflects documented differences in the geochemical environment. Analyses of sampled colloids included SEM and ICP-MS analysis of colloids on membrane and single particle analysis of samples in bottles. The colloid concentration was rather low in these Na-Mg-Ca-HCO 3 type waters. Trace element results show that transition metals and some heavy metals are associated with the colloid phase. Distribution coefficients of trace elements between the water and colloid phases were estimated. For example for uranium, an average of 200 pg/ml was detected in the water, and 40 pg/ml was detected in the colloid phase. A K p value of 2* 10 6 ml/g was calculated, considering (colloid) = 100 ng/ml. Groundwater samples were collected for analysis of the concentration of organic carbon (TOC), humic substances and metals associated with the humic substances. TOC varied in the range 4-14 mg/l in three boreholes, one borehole had a TOC<1.5 mg/l. The metal speciation study indicated that a large fraction, 8-67% of uranium was bound to the humic matter compared to the fractions of Ca and Fe (<0.4% and 0.02-10%, resp.). 60 refs, 8 figs, 16 tabs

  12. Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

    International Nuclear Information System (INIS)

    Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

    1996-01-01

    The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

  13. A novel start-up procedure for natural-circulation boiling water reactors

    International Nuclear Information System (INIS)

    Annalisa Manera; Frank Schaefer

    2005-01-01

    Full text of publication follows: The elimination of recirculation pumps and associated systems, as proposed for natural-circulation Boiling Water Reactors (BWRs), allow a great simplification in the design of BWRs. On the other hand, it has been shown both experimentally and analytically that such a new reactor configuration makes the system susceptible to thermal-hydraulic instabilities during the start-up phase (so-called flashing-induced instabilities). Therefore, appropriate start-up procedures have to be planned to avoid instabilities in natural-circulation BWRs. Not many proposals of start-up procedures for natural-circulation BWRs are reported in literature, but all authors agree on the fact that the system should be pressurized before the transition to two-phase circulation is allowed. Nayak [1] and Jiang and coauthors [2] proposed to externally pressurize the system by injecting in the pressure vessel respectively steam produced in a separate boiler or nitrogen. Once the pressure in the reactor vessel is high enough, the reactor power can be increased to achieve two-phase natural circulation. Unfortunately, the procedure suggested by Nayak requires an external boiler of adequate volume and power and the related connecting piping to the reactor vessel, while the procedure suggested by Jiang and coauthors requires an additional system for the nitrogen storage and the related connecting piping to the reactor vessel. The external pressurization does not accomplish to the requirements of simplicity that are at the very base of natural circulation BWRs design and it is thus not recommendable. Cheung and Rao [3] suggested a start-up procedure in which the reactor is first filled with water at 80 deg. C at a pressure of 0.55 bar. The reactor is made critical and is pressurized in conditions of single-phase circulation up to a pressure of 63 bar. At this pressure a sudden transition to two-phase operation is achieved by opening the MSIVs (Main Steam Isolation

  14. Experimental study on the safety of Kyoto University Research Reactor at natural circulation cooling mode

    International Nuclear Information System (INIS)

    Zhang, Jian; Shen, Xiuzhong; Fujihara, Yasuyuki; Sano, Tadafumi; Yamamoto, Toshihiro; Nakajima, Ken

    2015-01-01

    Highlights: • The natural circulation cooling capacity of Kyoto University Research Reactor (KUR) was experimentally investigated. • The distributions of the outlet temperature of the fuel elements under natural circulation operations were measured. • The average temperature rise and the average natural circulation flow velocity in core were calculated. • The safety of KUR under all of the normal operations with natural circulation cooling mode has been analyzed. • The natural circulation flow after the reactor shutdown was confirmed. - Abstract: In this study, the natural circulation cooling capacity of Kyoto University Research Reactor (KUR) is experimentally investigated by measuring the inlet and outlet temperatures of the core under natural circulation operation at various thermal powers ranging from 10 kW to 100 kW and the shutdown state. In view of the uneven power distribution and the resultant inconsistent coolant outlet temperature in the core, eight measuring points located separately in the outlet of the fuel elements were chosen to investigate the distribution of the outlet temperature of the core. The natural circulation cooling capacity represented by the average natural circulation flow velocity in the core is calculated from the temperature difference between the outlet and inlet temperature of the core. The measured outlet temperature of the fuel elements shows a cross-sectional distribution agreeing with the distribution of the thermal output of the fuel elements in the core. Since the measured outlet temperatures decrease quickly in the flow direction in a small local region above the outlet of the core, the mixing of the hot water out of the core with the cold water around the core outlet is found to happen in the small region not more than 5 cm far from the core outlet. The natural circulation flow velocity in the core increases non-linearly with the thermal power. The safety of KUR has been analysed by conservatively estimating the

  15. Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

    International Nuclear Information System (INIS)

    Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

    1987-01-01

    Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

  16. Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

    International Nuclear Information System (INIS)

    Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

    2015-01-01

    Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

  17. Neutronic design for a 100MWth Small modular natural circulation lead or lead-alloy cooled fast reactors core

    International Nuclear Information System (INIS)

    Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q.

    2015-01-01

    Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW th natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

  18. Influence of reactor design on the establishment of natural circulation in pool-type LMFBR

    International Nuclear Information System (INIS)

    Durham, M.E.

    1976-01-01

    The general principles involved in establishing natural circulation in a pool-type liquid metal cooled fast breeder reactor following loss of a.c. supplies are elucidated and the effects of design features by use of the computer code MELANI are quantified. It is shown that natural circulation can provide a feasible means of emergency core cooling in addition to that provided by pony motors. The choice of primary pump rundown time has a significant effect in controlling peak core outlet temperatures in the hypothetical case of natural circulation alone being the core heat removal process. (author)

  19. 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.

  20. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der.

    1989-01-01

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  1. Fractal reactor: An alternative nuclear fusion system based on nature's geometry

    International Nuclear Information System (INIS)

    Siler, T. L.

    2007-01-01

    The author presents his concept of the Fractal Reactor, which explores the possibility of building a plasma fusion power reactor based on the real geometry of nature [fractals], rather than the virtual geometry that Euclid postulated around 330 BC; nearly every architect of our plasma fusion devices has been influenced by his three-dimensional geometry. The idealized points, lines, planes, and spheres of this classical geometry continue to be used to represent the natural world and to describe the properties of all geometrical objects, even though they neither accurately nor fully convey nature's structures and processes. The Fractal Reactor concept contrasts the current containment mechanisms of both magnetic and inertial containment systems for confining and heating plasmas. All of these systems are based on Euclidean geometry and use geometrical designs that, ultimately, are inconsistent with the Non-Euclidean geometry and irregular, fractal forms of nature (3). The author explores his premise that a controlled, thermonuclear fusion energy system might be more effective if it more closely embodies the physics of a star

  2. Parametric study of natural circulation flow in molten salt fuel in molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pauzi, Anas Muhamad, E-mail: Anas@uniten.edu.my [Centre of Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Cioncolini, Andrea; Iacovides, Hector [School of Mechanical, Aerospace, and Civil Engineering (MACE), University of Manchester, Oxford Road, M13 9PL Manchester (United Kingdom)

    2015-04-29

    The Molten Salt Reactor (MSR) is one of the most promising system proposed by Generation IV Forum (GIF) for future nuclear reactor systems. Advantages of the MSR are significantly larger compared to other reactor system, and is mainly achieved from its liquid nature of fuel and coolant. Further improvement to this system, which is a natural circulating molten fuel salt inside its tube in the reactor core is proposed, to achieve advantages of reducing and simplifying the MSR design proposed by GIF. Thermal hydraulic analysis on the proposed system was completed using a commercial computation fluid dynamics (CFD) software called FLUENT by ANSYS Inc. An understanding on theory behind this unique natural circulation flow inside the tube caused by fission heat generated in molten fuel salt and tube cooling was briefly introduced. Currently, no commercial CFD software could perfectly simulate natural circulation flow, hence, modeling this flow problem in FLUENT is introduced and analyzed to obtain best simulation results. Results obtained demonstrate the existence of periodical transient nature of flow problem, hence improvements in tube design is proposed based on the analysis on temperature and velocity profile. Results show that the proposed system could operate at up to 750MW core power, given that turbulence are enhanced throughout flow region, and precise molten fuel salt physical properties could be defined. At the request of the authors and the Proceedings Editor the name of the co-author Andrea Cioncolini was corrected from Andrea Coincolini. The same name correction was made in the Acknowledgement section on page 030004-10 and in reference number 4. The updated article was published on 11 May 2015.

  3. Development of natural circulation small and medium sized boiling water reactor: HSBWR-600

    International Nuclear Information System (INIS)

    Miki, Minoru; Horiuchi, Tetsuo; Yoshimoto, Yuichiro; Sumida, Isao; Murase, Michio; Akita, Minoru; Niino, Tsuyoshi

    1988-01-01

    In nuclear power generation, the development of large reactors has been promoted as the main energy source in Japan. However, world economy entered low growth age, and the growth of electric power demand slowed down. Accordingly, attention has been paid to the medium and small reactors that can cope with whatever needs by serializing their types in addition to the nuclear power plants of medium output matching to electric power demand. In order to cope with these new needs, the economical efficiency of medium and small reactors must be as close as possible to that of large reactors, and as the countermeasures to the demerits due to small size, those must be made into the plants having simplified systems and the safety easily acceptable to public. Hitachi Ltd. plans to develop the natural circulation type medium and small BWRs of 600 NWe output class, HSBWR-600, on the basis of the nuclear power plant technology based on the rich results of design and operation of BWRs obtained so far, and to rank them as one of the BWR series. The target of their development design, the circumstance of their development, the core design and the thermo-hydraulic characteristics, the reactor pressure vessel and in-core structures, the safety design, system design, building layout and the evaluation are reported. (Kako, I.)

  4. Thermo-fluid analysis of water cooled research reactors in natural convection

    International Nuclear Information System (INIS)

    Veloso, Maria Auxiliadora Fortini

    2004-01-01

    The STHIRP-1 computer program, which fundamentals are described in this work, uses the principles of the subchannels analysis and has the capacity to simulate, under steady state and transient conditions, the thermal and hydraulic phenomena which occur inside the core of a water-refrigerated research reactor under a natural convection regime. The models and empirical correlations necessary to describe the flow phenomena which can not be described by theoretical relations were selected according to the characteristics of the reactor operation. Although the primary objective is the calculation of research reactors, the formulation used to describe the fluid flow and the thermal conduction in the heater elements is sufficiently generalized to extend the use of the program for applications in power reactors and other thermal systems with the same features represented by the program formulations. To demonstrate the analytical capacity of STHIRP-l, there were made comparisons between the results calculated and measured in the research reactor TRIGA IPR-R1 of CDTN/CNEN. The comparisons indicate that the program reproduces the experimental data with good precision. Nevertheless, in the future there must be used more consistent experimental data to corroborate the validation of the program. (author)

  5. Core Power Limits For A Lead-Bismuth Natural Circulation Actinide Burner Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee; Kim, D.; Todreas, N. E.; Mujid S. Kazimi

    2002-04-01

    The Idaho National Engineering and Environmental Laboratory and Massachusetts Institute of Technology are investigating the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The design being considered here is a pool type reactor that burns actinides and utilizes natural circulation of the primary coolant, a conventional steam power conversion cycle, and a passive decay heat removal system. Thermal-hydraulic evaluations of the actinide burner reactor were performed to determine allowable core power ratings that maintain cladding temperatures below corrosion-established temperature limits during normal operation and following a loss-of-feedwater transient. An economic evaluation was performed to optimize various design parameters by minimizing capital cost. The transient power limit was initially much more restrictive than the steady-state limit. However, enhancements to the reactor vessel auxiliary cooling system for transient decay heat removal resulted in an increased power limit of 1040 MWt, which was close to the steady-state limit. An economic evaluation was performed to estimate the capital cost of the reactor and its sensitivity to the transient power limit. For the 1040 MWt power level, the capital cost estimate was 49 mills per kWhe based on 1999 dollars.

  6. Nuclear power for coexistence with nature, high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

    Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

  7. Natural vibrations of a core banel of a PWR type reactor by elements of revolution shell

    International Nuclear Information System (INIS)

    Barcellos, C.S. de.

    1980-01-01

    Aim to estimate the behavior of the cove barrel of PWR type reactors, submitted to several load conditions, their dynamic characteristic, were determined. In order to obtain the natural modes and frequencies of the core barrel, the CYLDYFE comprete code based in the finite element method, was developed. The obtained results are compared with results obtained by other programs such as SAP, ASKA and STRUDL/DYNAL and by other analytical methods. (M.C.K.) [pt

  8. Removal of Iron and Manganese from Natural Groundwater by Continuous Reactor Using Activated and Natural Mordenite Mineral Adsorption

    Science.gov (United States)

    Zevi, Y.; Dewita, S.; Aghasa, A.; Dwinandha, D.

    2018-01-01

    Mordenite minerals derived from Sukabumi natural green stone founded in Indonesia was tested in order to remove iron and manganese from natural groundwater. This research used two types of adsorbents which were consisted of physically activated and natural mordenite. Physical activation of the mordenite was carried out by heating at 400-600°C for two hours. Batch system experiments was also conducted as a preliminary experiment. Batch system proved that both activated and natural mordenite minerals were capable of reducing iron and manganese concentration from natural groundwater. Then, continuous experiment was conducted using down-flow system with 45 ml/minute of constant flow rate. The iron & manganese removal efficiency using continuous reactor for physically activated and natural mordenite were 1.38-1.99%/minute & 0.8-1.49%/minute and 2.26%/minute & 1.37-2.26%/minute respectively. In addition, the regeneration treatment using NH4Cl solution managed to improve the removal efficiency of iron & manganese to 1.98%/minute & 1.77-1.90%/minute and 2.25%/minute & 2.02-2.21%/minute on physically activated mordenite and natural mordenite respectively. Subsequently, the activation of the new mordenite was carried out by immersing mordenite in NH4Cl solution. This chemical activation showed 2.42-2.75%/minute & 0.96 - 2.67 %/minute and 2.66 - 2.78 %/minute & 1.34 - 2.32 %/minute of iron & manganese removal efficiency per detention time for chemically activated and natural mordenite respectively.

  9. Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor

    International Nuclear Information System (INIS)

    Mohora, Emilijan; Rončević, Srdjan; Dalmacija, Božo; Agbaba, Jasmina; Watson, Malcolm; Karlović, Elvira; Dalmacija, Milena

    2012-01-01

    Highlights: ► A continuous electrocoagulation/flotation reactor was designed built and operated. ► Highest NOM removal according to UV 254 was 77% relative to raw groundwater. ► Highest NOM removal accordance to DOC was 71%, relative to raw groundwater. ► Highest As removal archived was 85% (6.2 μg/l), relative to raw groundwater. ► Specific reactor energy and electrode consumption was 1.7 kWh/m 3 and 66 g Al/m 3 . - Abstract: The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate = 4.3 l/h, inter electrode distance = 2.8 cm, current density = 5.78 mA/cm 2 , A/V ratio = 0.248 cm −1 . The NOM removal according to UV 254 absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 μg As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m 3 . According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater.

  10. Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mohora, Emilijan, E-mail: emohora@ifc.org [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia); Roncevic, Srdjan; Dalmacija, Bozo; Agbaba, Jasmina; Watson, Malcolm; Karlovic, Elvira; Dalmacija, Milena [University of Novi Sad Faculty of Sciences, Department of Chemistry, Biochemistry and Environmental Protection, Trg D. Obradovica 3, 21000 Novi Sad (Serbia)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer A continuous electrocoagulation/flotation reactor was designed built and operated. Black-Right-Pointing-Pointer Highest NOM removal according to UV{sub 254} was 77% relative to raw groundwater. Black-Right-Pointing-Pointer Highest NOM removal accordance to DOC was 71%, relative to raw groundwater. Black-Right-Pointing-Pointer Highest As removal archived was 85% (6.2 {mu}g/l), relative to raw groundwater. Black-Right-Pointing-Pointer Specific reactor energy and electrode consumption was 1.7 kWh/m{sup 3} and 66 g Al/m{sup 3}. - Abstract: The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate = 4.3 l/h, inter electrode distance = 2.8 cm, current density = 5.78 mA/cm{sup 2}, A/V ratio = 0.248 cm{sup -1}. The NOM removal according to UV{sub 254} absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 {mu}g As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m{sup 3}. According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater.

  11. On Stability of Natural-circulation-cooled Boiling Water Reactors during Start-up (Experimental Results)

    International Nuclear Information System (INIS)

    Manera, A.; Van der Hagen, T.H.J.J.

    2002-01-01

    The characteristics of flashing-induced instabilities, which are of importance during the start-up phase of natural-circulation Boiling Water Reactors (BWRs), are studied. Experiments at typical start-up conditions (low power and low pressure) are carried out on a steam/water natural circulation loop. The mechanism of flashing-induced instability is analyzed in detail and it is found that non-equilibrium between phases and enthalpy transport plays an important role in the instability process. Pressure and steam volume in the steam dome are found to have a stabilizing effect. The main characteristics of the instabilities have been analyzed. (authors)

  12. Oxygen concentration diffusion analysis of lead-bismuth-cooled, natural-circulation reactor

    International Nuclear Information System (INIS)

    Ito, Kei; Sakai, Takaaki

    2001-11-01

    The feasibility study on fast breeder reactors in Japan has been conducted at JNC and related organizations. The Phase-I study has finished in March, 2001. During the Phase-I activity, lead-bismuth eutectic coolant has been selected as one of the possible coolant options and a medium-scale plant, cooled by a lead-bismuth natural circulation flow was studied. On the other side, it is known that lead-bismuth eutectic has a problem of structural material corrosiveness. It was found that oxygen concentration control in the eutectic plays an important role on the corrosion protection. In this report, we have developed a concentration diffusion analysis code (COCOA: COncentration COntrol Analysis code) in order to carry out the oxygen concentration control analysis. This code solves a two-dimensional concentration diffusion equation by the finite differential method. It is possible to simulate reaction of oxygen and hydrogen by the code. We verified the basic performance of the code and carried out oxygen concentration diffusion analysis for the case of an oxygen increase by a refueling process in the natural circulation reactor. In addition, characteristics of the oxygen control system was discussed for a different type of the control system as well. It is concluded that the COCOA code can simulate diffusion of oxygen concentration in the reactor. By the analysis of a natural circulation medium-scale reactor, we make clear that the ON-OFF control and PID control can well control oxygen concentration by choosing an appropriate concentration measurement point. In addition, even when a trouble occurs in the oxygen emission or hydrogen emission system, it observes that control characteristic drops away. It is still possible, however, to control oxygen concentration in such case. (author)

  13. Removal of natural organic matter and arsenic from water by electrocoagulation/flotation continuous flow reactor.

    Science.gov (United States)

    Mohora, Emilijan; Rončević, Srdjan; Dalmacija, Božo; Agbaba, Jasmina; Watson, Malcolm; Karlović, Elvira; Dalmacija, Milena

    2012-10-15

    The performance of the laboratory scale electrocoagulation/flotation (ECF) reactor in removing high concentrations of natural organic matter (NOM) and arsenic from groundwater was analyzed in this study. An ECF reactor with bipolar plate aluminum electrodes was operated in the horizontal continuous flow mode. Electrochemical and flow variables were optimized to examine ECF reactor contaminants removal efficiency. The optimum conditions for the process were identified as groundwater initial pH 5, flow rate=4.3 l/h, inter electrode distance=2.8 cm, current density=5.78 mA/cm(2), A/V ratio=0.248 cm(-1). The NOM removal according to UV(254) absorbance and dissolved organic matter (DOC) reached highest values of 77% and 71% respectively, relative to the raw groundwater. Arsenic removal was 85% (6.2 μg As/l) relative to raw groundwater, satisfying the drinking water standards. The specific reactor electrical energy consumption was 17.5 kWh/kg Al. The specific aluminum electrode consumption was 66 g Al/m(3). According to the obtained results, ECF in horizontal continuous flow mode is an energy efficient process to remove NOM and arsenic from groundwater. Copyright © 2012 Elsevier B.V. All rights reserved.

  14. Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

    International Nuclear Information System (INIS)

    Kamide, H.; Kimura, N.; Miyakoshi, H.; Nagasawa, K.

    2001-01-01

    Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

  15. Power optimization in the star-LM modular, natural convection reactor system

    International Nuclear Information System (INIS)

    Spencer, B.W.; Sienicki, J.J.; Farmer, M.T.

    2001-01-01

    The secure, transportable, autonomous reactor (STAR) project addresses the needs of developing countries and independent power producers for small, multi-purpose energy systems, which operate near autonomously for very long term. The STAR-LM variant described here is a liquid metal cooled, fast reactor system. Previous development of STAR-LM resulted in a 300 MWt modular, pool-type reactor based on criteria for factory fabrication, full transportability (barge, overland, rail), and fast construction and startup. Steam generator modules are placed directly into the primary heat transport circuit, eliminating the intermediate heat transport loop. Natural convection heat transport at all power levels eliminates the need for main coolant pumps. Seismic isolation eliminates concern about seismic and sloshing-related loads in the pool configuration. Even end-of-spectrum postulated events such as loss-of-heat sink with failure to scram are terminated passively by inherent core power shutdown, and decay heat is passively rejected to the atmospheric air inexhaustible heat sink by guard vessel exterior cooling. Recent concept development has focused on maximizing the power achievable in a small module size based on preserving key criteria for: full spectrum of modes of module transport from factory to site (including rail transport); 100% natural circulation heat transport; ultra-long core cartridge lifetime; coolant and cladding peak temperatures well within the existing (Russian) database for Pb/Bi coolant and ferritic steel core materials. (author)

  16. Temperature control characteristics analysis of lead-cooled fast reactor with natural circulation

    International Nuclear Information System (INIS)

    Yang, Minghan; Song, Yong; Wang, Jianye; Xu, Peng; Zhang, Guangyu

    2016-01-01

    Highlights: • The LFR temperature control system are analyzed with frequency domain method. • The temperature control compensator is designed according to the frequency analysis. • Dynamic simulation is performed by SIMULINK and RELAP5-HD. - Abstract: Lead-cooled Fast Reactor (LFR) with natural circulation in primary system is among the highlights in advance nuclear reactor research, due to its great superiority in reactor safety and reliability. In this work, a transfer function matrix describing coolant temperature dynamic process, obtained by Laplace transform of the one-dimensional system dynamic model is developed in order to investigate the temperature control characteristics of LFR. Based on the transfer function matrix, a close-loop coolant temperature control system without compensator is built. The frequency domain analysis indicates that the stability and steady-state of the temperature control system needs to be improved. Accordingly, a temperature compensator based on Proportion–Integration and feed-forward is designed. The dynamic simulation of the whole system with the temperature compensator for core power step change is performed with SIMULINK and RELAP5-HD. The result shows that the temperature compensator can provide superior coolant temperature control capabilities in LFR with natural circulation due to the efficiency of the frequency domain analysis method.

  17. Feasibility analysis of the Primary Loop of Pool-Type Natural Circulating Nuclear Reactor Dedicated to Seawater Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woonho; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, the feasibility of natural circulation was evaluated for the reference plant AHR400 (Advanced Heating Reactor 400MWth). AHR400 is a pool-type desalination-dedicated nuclear reactor. As a consequence, AHR400 has low operating pressure and temperature which provides large safety margin. Removal of the reactor coolant pump from the AHR400 will enforce integrity of the reactor vessel and passive safety feature. Therefore, the study also tried to find out optimized primary loop design to achieve total natural circulation of the coolant. Natural circulation capacity of the primary loop of the desalination dedicated nuclear reactor AHR400 was evaluated. It was concluded that to remove RCP from the AHR400 and operates the reactor only by natural circulation of the coolant is impossible. Decreased core power as half make removal of RCP possible with 15m central height difference between the core and IHXs. Furthermore, validation and modification of pressure loss coefficients by small-scaled natural circulation experiment at a pool-type reactor would provide more accurate results.

  18. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  19. Steam drum level control studies of a natural circulation multi loop reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

    2013-12-15

    The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

  20. Natural Circulation Level Optimization and the Effect during ULOF Accident in the SPINNOR Reactors

    International Nuclear Information System (INIS)

    Abdullah, Ade Gafar; Su'ud, Zaki; Kurniadi, Rizal; Kurniasih, Neny; Yulianti, Yanti

    2010-01-01

    Natural circulation level optimization and the effect during loss of flow accident in the 250 MWt MOX fuelled small Pb-Bi Cooled non-refueling nuclear reactors (SPINNOR) have been performed. The simulation was performed using FI-ITB safety code which has been developed in ITB. The simulation begins with steady state calculation of neutron flux, power distribution and temperature distribution across the core, hot pool and cool pool, and also steam generator. When the accident is started due to the loss of pumping power the power distribution and the temperature distribution of core, hot pool and cool pool, and steam generator change. Then the feedback reactivity calculation is conducted, followed by kinetic calculation. The process is repeated until the optimum power distribution is achieved. The results show that the SPINNOR reactor has inherent safety capability against this accident.

  1. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  2. CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Haden, T.H.J.J. van der; Zboray, R.; Manera, A.; Mudde, R.F.

    2002-01-01

    At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

  3. Investigation of the transition from forced to natural convection in the research reactor Munich II

    International Nuclear Information System (INIS)

    Skreba, S.; Adamek, J.; Unger, H.

    1999-01-01

    The new research reactor Munich II (FRM-II), which is under construction at the Technical University Munich, Germany, makes use of a newly developed compact reactor core consisting of a single fuel element, which is assembled of two concentric pipes. Between the fuel element's inner and outer pipe 113 involutely bent fuel plates are placed rotationally symmetric, forming 113 cooling channels of a constant width of 2.2 mm. After a shut down of the reactor, battery supported cooling pumps are started by the reactor safety system in order to remove the decay heat by a downwards directed forced flow. Three hours after they have been started, the cooling pumps are shut down and so-called 'natural convection flaps' are opened by their own weight. Through a flow path, which is provided by the opening of the natural convection flaps, the decay heat is given off to the water in the reactor pool after the direction of the flow has changed and an upwards directed natural convection flow has developed. At the Department for Nuclear and New Energy Systems of the Ruhr-University Bochum, Germany, a test facility has been built in order to confirm the concept of the decay heat removal in the FRM-II, to acquire data of single and two phase natural convection flows and to detect the dry out in a narrow channel. The thermohydraulics of the FRM-II are simulated by an electrically heated test section, which represents one cooling channel of the fuel element. At first experiments have been performed, which simulated the transition from forced to natural convection in the core of the FRM-II, both at normal operation and at a complete loss of the decay heat removal pumps. In case of normal operation, the transition from forced to natural convection takes place single phased. If a complete loss of the active decay heat removal system occurs, the decay heat removal is ensured by a quasi-steady two phase flow. In a second test series minimum heat flux densities leading to pressure pulsations

  4. A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

    International Nuclear Information System (INIS)

    Baltazar, O.

    1978-01-01

    A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt

  5. The use of natural analogues in the long-term extrapolation of glass corrosion processes

    International Nuclear Information System (INIS)

    Lutze, W.; Grambow, B.; Ewing, R.C.; Jercinovic, M.J.

    1987-01-01

    One of the most critical aspects of nuclear waste management is the extrapolation of materials and systems behavior from short term experiments, typically on the order of one year, over comparatively very long periods of time. Safety and risk analyses have to rely on extrapolations and the respective findings have to be evaluated in the frame of licensing procedures. In this unique situation, any source of information that can lend support to the credibility of predicted behavior, should be exploited and investigated with great care. There are natural systems, e.g. the Oklo reactor, which can provide evidence of radionuclide migration over very long periods of time and thus help to answer specific questions of interest. Natural glasses and minerals can serve as analogues for both glass and crystalline nuclear waste forms, and the alteration of the natural materials can be studied to infer information on the behavior of the man-made products in geologic environments. This paper reviews most of the work performed by the authors and their colleagues in this field together with information available from literature and discusses the extent to which natural glasses can be used to validate or verify predictions. (author)

  6. Study of core flow distribution for small modular natural circulation lead or lead-alloy cooled fast reactors

    International Nuclear Information System (INIS)

    Chen, Zhao; Zhao, Pengcheng; Zhou, Guangming; Chen, Hongli

    2014-01-01

    Highlights: • A core flow distribution calculation code for natural circulation LFRs was developed. • The comparison study between the channel method and the CFD method was conducted. • The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted. - Abstract: Small modular natural circulation lead or lead-alloy cooled fast reactor (LFR) is a potential candidate for LFR development. It has many attractive advantages such as reduced capital costs and inherent safety. The core flow distribution calculation is an important issue for nuclear reactor design, which will provide important input parameters to thermal-hydraulic analysis and safety analysis. The core flow distribution calculation of a natural circulation LFR is different from that of a forced circulation reactor. In a forced circulation reactor, the core flow distribution can be controlled and adjusted by the pump power and the flow distributor, while in a natural circulation reactor, the core flow distribution is automatically adjusted according to the relationship between the local power and the local resistance feature. In this paper, a non-uniform heated parallel channel flow distribution calculation code was developed and the comparison study between the channel method and the CFD method was carried out to assess the exactness of the developed code. The core flow distribution analysis and optimization design for a 10MW natural circulation LFR was conducted using the developed code. A core flow distribution optimization design scheme for a 10MW natural circulation LFR was proposed according to the optimization analysis results

  7. Natural circulation of integrated-type marine reactor at inclined attitude

    International Nuclear Information System (INIS)

    Iyori, Isao; Aya, Izuo; Murata, Hiroyuki; Kobayashi, Michiyuki; Nariai, Hideki

    1987-01-01

    A steady-state single-phase natural circulation test was performed to clarify the effect of inclination by using a model of an integrated-type marine reactor. It was found that several types of flow pattern occur in the natural circulation loop corresponding to the range of inclination angle. Stable flow rates are sustained up to near 90 0 because of the occurrence of a driving force arising from those sections of the facility which were horizontal before the inclination. It was found that the temperature distribution in the steam generator at inclined attitude depends essentially only on the elevation z. The applicability of a one-dimensional analytical model was examined. It was clarified that employment of detailed U-turn flow paths, their correlation, and temperature-distribution function of core is essential for improvement. (orig.)

  8. Operating experience of natural circulation core cooling in boiling water reactors

    International Nuclear Information System (INIS)

    Kullberg, C.; Jones, K.; Heath, C.

    1993-01-01

    General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

  9. Design of an additional heat sink based on natural circulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Frischengruber, Kurt; Solanilla, Roberto; Fernandez, Ricardo; Blumenkrantz, Arnaldo; Castano, Jorge

    1989-01-01

    Residual heat removal through the steam generators in Nuclear Power Plant with pressurized water reactors (PWR) or pressurized heavy water reactors (PHWR in pressured vessel or pressured tube types) requires the maintenance of the steam generator inventory and the availability of and appropriate heat sink, which are based on the operability of the steam generators feedwater system. This paper describes the conceptual design of an assured heat removal system which includes only passive elements and is based on natural circulation. The system can supplement the original systems of the plant. The new system includes a condenser/boiler heat exchanger to condense the steam produced in the steam generator, transferring the heat to the water of an open pool at atmospheric pressure. The condensed steam flows back to the steam generators by natural circulation effects. The performance of an Atucha type PHWR nuclear power station with and without the proposed system is calculated in an emergency power case for the first 5000 seconds after the incident. The analysis shows that the proposed system offers the possibility to cool-down the plant to a low energy state during several hours and avoids the repeated actuation of the primary and secondary system safety valves. (Author) [es

  10. Simulation of natural convection cooling phenomena for research reactors using the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Al-Habit, E.

    2006-01-01

    This study deals with testing the capacity of the code PARET to simulate natural circulation phenomena under different boundary conditions in addition to assessment of some new options related to simulation of control rod movement and the reactivity effect of thermal expansion fuel elements. the experiments of the simple thermal hydraulic loop of Missouri University about natural circulation phenomena in narrow parallel channel were used to validate the code. The results indicate good agreements regarding the evolution of coolant velocity and clad temperature. In particular the heat transfer coefficient of natural convection has been calculated in good agreement with the experiment. On the other hand, the core of MNSR reactor has been modelled to stimulate the reactor dynamic behaviour under natural circulation condition for different initial power level. The observed oscillations during the initial phase vanish gradually with passing time. In this context three experiment of step reactivity insertion were calculated using two different options of boundary conditions, either using initial velocity or pressure drop along the core. The results indicate good agreement with the experiments regarding the evolution of relative power. The validations included also sensitivity analysis against some important parameters like initial velocity and radial distance of fuel rod. The new option for simulation of control rod movement was also tested. For this purpose the MNSR experiment of all control rod withdraw was selected. This means control rod velocity was estimated using experimental measurement. The simulation result of relative power evolution shows good agreement with the experiment during the first phase of the transient. However, an increased deviation is observed in the following phase due to the effect of closed hydrodynamics loop, which can be modelled with the code PARET. (Authors)

  11. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

    2016-11-30

    The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

  12. Application of natural adsorbents as decontamination agents for the elimination of the consequences of the Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Tarasevich, Yu.I.

    1996-01-01

    The scientific foundations of using natural adsorbents as ion exchangers,filtering media and adagulants for water purification ase presented. The results showing the efficiency of practical application of natural adsorbents for the decontamination of water, clothes, machinery, construction materials, etc. during the elimination of the consequences of the Chernobyl reactor accident in 1986-1987 are presented

  13. Natural circulation analysis for the advanced neutron source reactor refueling process 11

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, R.F.; Dasardhi, S.; Elkassabgi, Y. [Texas A& M Univ., Kingsville, TX (United States); Yoder, G.L. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    During the refueling process of the Advanced Neutron Source Reactor (ANSR), the spent fuel elements must be moved from the primary coolant loop (containing D{sub 2}O), through a heavy water pool, and finally into a light water spent fuel storage area. The present refueling scheme utilizes remote refueling equipment to move the spent fuel elements through a D{sub 2}O filled stack and tunnel into a temporary storage canal. A transfer lock is used to move the spent fuel elements from the D{sub 2}O-filled interim storage canal to a light water pool. Each spent fuel element must be cooled during this process, using either natural circulation or forced convection. This paper presents a summary of the numerical techniques used to analyze natural circulation cooling of the ANSR fuel elements as well as selected results of the calculations. Details of the analysis indicate that coolant velocities below 10 cm/s exist in the coolant channels under single phase natural circulation conditions. Also, boiling does not occur within the channels if power levels are below a few hundred kW when the core transitions to natural circulation conditions.

  14. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  15. Studies on natural circulation cooling enhancement in a spent fuel in fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maekawa, Isamu; Akamatsu, Mikio; Toda, Shinichi; Sato, Manabu [Kawasaki Heavy Industries Ltd., Kobe (Japan); Mayumi, Masami

    2001-01-01

    Fast breeder reactor (FBR) has some advantages such as effective application of plutonium, excellent capacity to fire minor-actinides (longer half-life nuclides such as Np, Am, Cm, and so on) contained in radioactive wastes in the reactor to convert their shorter half-life nuclides. However, fuels containing the minor-actinides have a characteristic with higher exotherm and radioactive intensity than those of conventional ones, it is essential at their actual stages to prepare some rational fuel handling systems on their transportation, storage and so forth. In addition, there are few examples on natural circulation heat transfer test of a liquid metal using long sized container. Then, in order to establish an evaluating method on decay-heat removing property of a spent fuel assembly in sodium canister and pot, some natural circulation tests on a long sized container including a quasi pin-bundle structure for a working fluid of lead-bismuth (Pb-Bi) mixture with easier handling than that of sodium was carried out. A specimen could be mounted at optional angles from horizontal to vertical positions so as to evaluate effects of inclined angles. In addition, in order to estimate temperature and flow rate distribution in a long sized container and understand thermal flowing phenomenon in specimen system, numerical analysis using multi-dimensional analysis code was carried out. As a result, it was found that in vertical arrangement system, natural circulation phenomenon is limited at upper portion of the exothermal portion, and its maximum temperature was tested at central portion of top pin-bundle of the exothermal portion. And, it was also found that at horizontal arrangement maximum temperature was 40 centigrade less than that of vertical arrangement, and so forth. (G.K.)

  16. The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

    International Nuclear Information System (INIS)

    Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

    1993-01-01

    Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

  17. Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

    International Nuclear Information System (INIS)

    Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

    2011-01-01

    Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

  18. Natural convection in closed vertical cylinders with particular reference to gas cooled reactor standpipes

    International Nuclear Information System (INIS)

    Spence, I.D.

    1975-09-01

    The access to the core for fuel assemblies and control rods of the Advanced Gas Cooled Reactor is through the top cap by means of standpipes. The standpipe is essentially a cylindrical, vertical tube with cooled side wall, closed upper end and an orifice at the lower end which is exposed to the hot core fluid. This creates confined natural convection flow in the empty standpipe and this is the subject of this thesis. The investigation is carried out using analytical and experimental methods. For the analytical work, solution of laminar and turbulent flow is attempted using finite-difference computer techniques. The laminar flow performance is evaluated using two different finite-difference procedures, and the results are compared to each other and to existing analytical and experimental results for the open thermosyphon with cool inflow and hot sidewall, i.e. the complementary problem to the present one. For turbulent flow a two equation turbulence model is employed which provides transport equations for the kinetic energy of turbulence and its dissipation rate. The experimental rig is a full scale replica of the Advanced Gas Cooled Reactor control rod mechanism standpipe. Carbon dioxide and helium are used as the working fluids for the series of tests. (author)

  19. Enhanced CANDU6: Reactor and fuel cycle options - Natural uranium and beyond

    International Nuclear Information System (INIS)

    Ovanes, M.; Chan, P. S. W.; Mao, J.; Alderson, N.; Hopwood, J. M.

    2012-01-01

    The Enhanced CANDU 6 R (ECo R ) is the updated version of the well established CANDU 6 family of units incorporating improved safety characteristics designed to meet or exceed Generation III nuclear power plant expectations. The EC6 retains the excellent neutron economy and fuel cycle flexibility that are inherent in the CANDU reactor design. The reference design is based on natural uranium fuel, but the EC6 is also able to utilize additional fuel options, including the use of Recovered Uranium (RU) and Thorium based fuels, without requiring major hardware upgrades to the existing control and safety systems. This paper outlines the major changes in the EC6 core design from the existing C6 design that significantly enhance the safety characteristics and operating efficiency of the reactor. The use of RU fuel as a transparent replacement fuel for the standard 37-el NU fuel, and several RU based advanced fuel designs that give significant improvements in fuel burnup and inherent safety characteristics are also discussed in the paper. In addition, the suitability of the EC6 to use MOX and related Pu-based fuels will also be discussed. (authors)

  20. Natural uranium fueled light water moderated breeding hybrid power reactors: a feasibility study

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    1978-06-01

    The first part of the study consists of a thorough investigation of the properties of subcritical thermal lattices for hybrid reactor applications. Light water is found to be the best moderator for (fuel-self-sufficient) FSS hybrid reactors for power generation. Several lattice geometries and compositions of particular promise for LWHRs are identified. Using one of these lattices, fueled with natural uranium, the performance of several concepts of LWHR blankets is investigated, and optimal blanket designs are identified. The effect of blanket coverage efficiency and the feasibility of separating the functions of tritium breeding and of power generation to different blankets are investigated. Optimal iron-water shields for LWHRs are also determined. The performance of generic types of LWHRs is evaluated. The evolution of the blanket properties with burnup is evaluated and fuel management schemes are briefly examined. The feasibility of using the lithium system of the blanket to control the blanket power amplitude and shape is also investigated. A parametric study of the energy balance of LWHR power plants is carried out, and performance parameters expected from LWHRs are estimated. Discussions are given of special features of LWHRs and their fuel cycle

  1. Technical report on natural evaporation system for radioactive liquid waste treatment arising from TRIGA research reactors' decontamination and decommissioning activities

    International Nuclear Information System (INIS)

    Moon, J. S.; Jung, K. J.; Baek, S. T.; Jung, U. S.; Park, S. K.; Jung, K. H.

    1999-01-01

    This technical report described that radioactive liquid waste treatment for dismantling/decontamination of TRIGA Mark research reactor in Seoul. That is, we try safety treatment of operation radioactive liquid waste during of operating TRIGA Mark research reactor and dismantling radioactive liquid waste during R and D of research reactor hereafter, and by utilizing of new natural evaporation facility with describing design criteria of new natural evaporation facility. Therefore, this technical report described the quantity of present radioactive liquid waste and dismantling radioactive liquid waste hereafter, analysis the status of radial-rays/radioactivity, and also treatment method of this radioactive liquid waste. Also, we derived the method that the safeguard of outskirts environment and the cost down of radioactive liquid waste treatment by minimize of the radioactive liquid waste quantities, through-out design/operation of new natural evaporation facility for treatment of operation radioactive liquid waste and dismantling radioactive liquid waste. (author). 6 refs., 12 tabs., 5 figs

  2. Investigation of efficient {sup 131}I production from natural uranium at Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khalafi, H. [Nuclear Research Center, AEOI, No. 54 North Kargar Avenue, P.O. Box 14155/1339, Tehran (Iran, Islamic Republic of)]. E-mail: hossein_khalafi@yahoo.com; Nazari, K. [Jaber-Ibne-Hayan Research laboratories, AEOI, P.O. Box 11365/8486, Tehran (Iran, Islamic Republic of); Ghannadi-Maragheh, M. [Jaber-Ibne-Hayan Research laboratories, AEOI, P.O. Box 11365/8486, Tehran (Iran, Islamic Republic of)

    2005-05-15

    Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of {sup 131}I production in Tehran research reactor (TRR) was studied by two different methods. Primarily, standard nuclear codes such as ORIGEN, WIMS and CITATION were applied and then analytical solutions technique was followed. Calculated results and experimental works in the bench scale indicate that, by irradiation of 100 g natural Uranium (UO{sub 2}) for 100 h at 3.5 x 10{sup 13} (n's/cm{sup 2} s) thermal neutron flux in the TRR, one can produce about 5 Ci of {sup 131}I for medical purposes, on the other hand can produce very useful radionuclides like {sup 99}Mo and {sup 133}Xe in one batch irradiation in the unique production line.

  3. Investigation of efficient 131I production from natural uranium at Tehran research reactor

    International Nuclear Information System (INIS)

    Khalafi, H.; Nazari, K.; Ghannadi-Maragheh, M.

    2005-01-01

    Iodine-131, which has a half-life of 8.05 days, is the one of the most widely used radionuclides in medical diagnosis and treats some diseases of thyroid gland. Optimization of 131 I production in Tehran research reactor (TRR) was studied by two different methods. Primarily, standard nuclear codes such as ORIGEN, WIMS and CITATION were applied and then analytical solutions technique was followed. Calculated results and experimental works in the bench scale indicate that, by irradiation of 100 g natural Uranium (UO 2 ) for 100 h at 3.5 x 10 13 (n's/cm 2 s) thermal neutron flux in the TRR, one can produce about 5 Ci of 131 I for medical purposes, on the other hand can produce very useful radionuclides like 99 Mo and 133 Xe in one batch irradiation in the unique production line

  4. Migration of U-series radionuclides around the Bangombe natural fission reactor (Gabon)

    International Nuclear Information System (INIS)

    Bros, R.; Yanase, N.; Isobe, H.; Sato, T.; Iida, Y.; Ohnuki, T.; Roos, P.; Holm, E.

    1999-01-01

    The Bangombe natural fission reactors has undergone extensive weathering phenomena and continues to be affected by the penetration of meteoric waters. Hence this system provides a model for studying the stability of spent fuel uraninite and the influence of various rock matrices on the mobilization/retardation of various actinides and fission products. The Bangombe uranium deposit has been investigated by drilling on a grid. Radiochemical analysis by alpha- and gamma-spectroscopy of the obtained rocks show significant disequilibria of the 234 U/ 238 U, 230 Th/ 234 U, and 226 Ra/ 230 Th parent-daughter pairs. In this paper, a conceptual model for spatio/temporal evolution of the Bangombe system is proposed. (J.P.N.)

  5. Environment report 1990 of the Federal Minister for the Environment, Nature Protection and Reactor Safety

    International Nuclear Information System (INIS)

    1990-01-01

    The 'Environment Report 1990' describes the environmental situation in the Federal Republic of Germany; draws a balance of environmental policy measures taken and introduced; gives information on future fields of action in environmental policy. The 'Environment Report 1990' also deals with the 'Environment Expert Opinion 1987', produced by the board of experts on environmental questions. It contains surveys of the following sectors: Protection against hazardous materials air pollution abatement, water management, waste management, nature protection and preservation of the countryside, soil conservation, noise abatement, radiation protection, reactor safety. A separate part of the 'Environment Report 1990' deals with the progress made in 'interdisciplinary fields' (general law on the protection of the environment, instruments of environmental policy, environmental information and environmental research, transfrontier environmental policy). (orig./HP) [de

  6. Experimental investigation of natural convection in a core of a marine reactor in rolling motion

    International Nuclear Information System (INIS)

    Murata, Hiroyuki; Sawada, Ken-ichi; Kobayashi, Michiyuki

    2000-01-01

    A series of single-phase natural circulation experiments in a simulated marine reactor mounted on a rolling bed was performed and the average Nusselt number in the core was evaluated in order to investigate effects of the rolling motion on the heat transfer in the core. Heat transfer with an upright attitude is well correlated with the Rayleigh number and is slightly lower than El-Genk's correlation. Heat transfer in the core is not affected by the inclination angle because the inclination of the present experiment is not large enough to cause any remarkable changes in the flow pattern of the core. Heat transfer in the core is enhanced by the rolling motion which is thought to cause internal flow in the core. Heat transfer during the rolling motion is correlated with the Richardson number for rolling motion, Ri R , and is classified into three regimes: (1) region A (0.05 R ≤0.3) where heat transfer is dominated by the inertial force due to the rolling motion; (2) region B (0.3 R ≤2) where heat transfer is affected by the combined effect of the inertial force and natural convection; and (3) region C (Ri R >2) where heat transfer is affected only by the natural convection. (author)

  7. Natural precursor based hydrothermal synthesis of sodium carbide for reactor applications

    Science.gov (United States)

    Swapna, M. S.; Saritha Devi, H. V.; Sebastian, Riya; Ambadas, G.; Sankararaman, S.

    2017-12-01

    Carbides are a class of materials with high mechanical strength and refractory nature which finds a wide range of applications in industries and nuclear reactors. The existing synthesis methods of all types of carbides have problems in terms of use of toxic chemical precursors, high-cost, etc. Sodium carbide (Na2C2) which is an alkali metal carbide is the least explored one and also that there is no report of low-cost and low-temperature synthesis of sodium carbide using the eco-friendly, easily available natural precursors. In the present work, we report a simple low-cost, non-toxic hydrothermal synthesis of refractory sodium carbide using the natural precursor—Pandanus. The formation of sodium carbide along with boron carbide is evidenced by the structural and morphological characterizations. The sample thus synthesized is subjected to field emission scanning electron microscopy (FESEM), x-ray powder diffraction (XRD), ultraviolet (UV)—visible spectroscopy, Fourier transform infrared spectroscopy (FTIR), Raman, and photoluminescent (PL) spectroscopic techniques.

  8. Parametric studies to establish natural circulation in advanced heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

    Design of Advanced Heavy Water Reactor (AHWR) is in progress. It consists of vertical pressure tubes with boiling light water coolant flowing through the tubes and heavy water moderator in the calandria. In PHWRs, core heat removal is through forced circulation of the coolant by PHT pumps. In AHWR, no PHT pumps are used and core heat is carried away by natural circulation of the coolant due to density difference between steam/water mixture inside the core and the water region outside the core. This passive means of core heat removal results in a number of benefits viz. (a) extra length of piping, valves, instruments, power supply and control systems for functioning of instruments are eliminated, (b) plant layout is simplified, (c) maintenance of valves and instruments is reduced. Natural circulation in AHWR is achieved by keeping the steam drum at a sufficient height above the core to get the required driving force. The loop height depends on many factors i.e. channel power, V{sub c}/V{sub f} ratio (ratio of coolant volume to fuel volume) and core height. The effect of these parameters on the loop height to establish natural circulation have been studied and presented. (author). 1 ref., 1 fig., 1 tab.

  9. Impact of design options on natural circulation performance of the AFR-300 advanced fast reactor

    International Nuclear Information System (INIS)

    Dunn, F. D.

    2002-01-01

    The AFR-300, Advanced Fast Reactor (300 Mwe), has been proposed as a Generation IV concept. It could also be used to dispose of surplus weapons grade plutonium or as an actinide burner for transmutation of high level radioactive waste. AFR-300 uses metallic fuel and sodium coolant. The design of AFR-300 takes account of the successful design and operation of EBR-II, but the AFR-300 design includes a number of advances such as an advanced fuel cycle, inspectability and improved economics. One significant difference between AFR-300 and EBR-II is that AFR-300 is considerably larger. Another significant difference is that AFR-300 has no auxiliary EM pump in the primary loop to guarantee positive core flow when the main primary pumps are shut down. Thus, one question that has come up in connection with the AFR-300 design is whether natural circulation flow is sufficient to prevent damage to the core if the primary pumps fail. Insufficient natural circulation flow through the core could result in high cladding temperatures and cladding failure due to eutectic penetration of the cladding by the metal fuel. The rate of eutectic penetration of the cladding is strongly temperature dependent, so cladding failure depends on how hot the cladding gets and how long it is at elevated temperatures. To investigate the adequacy of natural circulation flow, a number of pump failure transients and a number of design options have been analyzed with the SASSYS-1 systems analysis code. This code has been validated for natural circulation behavior by analysis of Shutdown Heat Removal Tests performed in EBR-II. The AFR-300 design includes flywheels on the primary pumps to extend the pump coastdown times, and the size of the flywheels can be picked to give optimum coastdown times. One series of transients that has been run consists of protected loss-of-flow transients with various values for the combined moment of inertia of the pump, the motor and the flywheel giving coastdown times from 70

  10. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  11. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France

    International Nuclear Information System (INIS)

    Gaussens, J.; Tanguy, P.

    1964-01-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  12. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Science.gov (United States)

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

  13. Research on enhancement of natural circulation capability in lead–bismuth alloy cooled reactor by using gas-lift pump

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Juanli, E-mail: Jenyzuo@163.com; Tian, Wenxi, E-mail: wxtian@mail.xjtu.edu.cn; Chen, Ronghua, E-mail: ronghua.chen@stu.xjtu.edu.cn; Qiu, Suizheng; Su, Guanghui, E-mail: ghsu@mail.xjtu.edu.cn

    2013-10-15

    Highlights: • The gas-lift pump has been adopted to enhance the natural circulation capability. • LENAC code is developed in my study. • The calculation results by LENAC code show good agreement with experiment results. • Gas mass flow rate, bubble diameter, rising pipe length are important parameters. -- Abstract: The gas-lift pump has been adopted to enhance the natural circulation capability in the type of lead–bismuth alloy cooled reactors such as Accelerator Driven System (ADS) and Liquid–metal Fast Reactor (LMFR). The natural circulation ability and the system safety are obviously influenced by the two phase flow characteristics of liquid metal–inert gas. In this study, LENAC (LEad bismuth alloy NAtural Circulation capability) code has been developed to evaluate the natural circulation capability of lead–bismuth cooled ADS with gas-lift pump. The drift flow theory, void fraction prediction model and friction pressure drop prediction model have been incorporated into LENAC code. The calculation results by LENAC code show good agreement with experiment results of CIRCulation Experiment (CIRCE) facility. The effects of the gas mass flow rate, void fraction, gas quality, bubble diameter and the rising pipe height or the potential difference between heat exchanger and reactor core on natural circulation capability of gas-lift pump have been analyzed. The results showed that in bubbly flow pattern, for a fixed value of gas mass flow rate, the natural circulation capability increased with the decrease of the bubble diameter. In the bubbly flow, slug flow, churn flow and annular flow pattern, with the gas mass flow rate increasing, the natural circulation capability initially increased and then declined. And the flow parameters influenced the thermal hydraulic characteristics of the reactor core significantly. The present work is helpful for revealing the law of enhancing the natural circulation capability by gas-lift pump, and providing theoretical

  14. Hydrogen safety risk assessment methodology applied to a fluidized bed membrane reactor for autothermal reforming of natural gas

    NARCIS (Netherlands)

    Psara, N.; Van Sint Annaland, M.; Gallucci, F.

    2015-01-01

    The scope of this paper is the development and implementation of a safety risk assessment methodology to highlight hazards potentially prevailing during autothermal reforming of natural gas for hydrogen production in a membrane reactor, as well as to reveal potential accidents related to hydrogen

  15. Thermo-fluid analysis of water cooled research reactors in natural convection; Analise termofluidodinamica de reatores nucleares de pesquisa refrigerados a agua em regime de conveccao natural

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Maria Auxiliadora Fortini

    2004-07-01

    The STHIRP-1 computer program, which fundamentals are described in this work, uses the principles of the subchannels analysis and has the capacity to simulate, under steady state and transient conditions, the thermal and hydraulic phenomena which occur inside the core of a water-refrigerated research reactor under a natural convection regime. The models and empirical correlations necessary to describe the flow phenomena which can not be described by theoretical relations were selected according to the characteristics of the reactor operation. Although the primary objective is the calculation of research reactors, the formulation used to describe the fluid flow and the thermal conduction in the heater elements is sufficiently generalized to extend the use of the program for applications in power reactors and other thermal systems with the same features represented by the program formulations. To demonstrate the analytical capacity of STHIRP-l, there were made comparisons between the results calculated and measured in the research reactor TRIGA IPR-R1 of CDTN/CNEN. The comparisons indicate that the program reproduces the experimental data with good precision. Nevertheless, in the future there must be used more consistent experimental data to corroborate the validation of the program. (author)

  16. Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  17. Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B. [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4

  18. On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

    International Nuclear Information System (INIS)

    Melesed'Hospital, G.

    1983-01-01

    The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

  19. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  20. Reliability analysis of 2400 MWth gas-cooled fast reactor natural circulation decay heat removal system

    International Nuclear Information System (INIS)

    Marques, M.; Bassi, C.; Bentivoglio, F.

    2012-01-01

    In support to a PSA (Probability Safety Assessment) performed at the design level on the 2400 MWth Gas-cooled Fast Reactor, the functional reliability of the decay heat removal system (DHR) working in natural circulation has been estimated in two transient situations corresponding to an 'aggravated' Loss of Flow Accident (LOFA) and a Loss of Coolant Accident (LOCA). The reliability analysis was based on the RMPS methodology. Reliability and global sensitivity analyses use uncertainty propagation by Monte Carlo techniques. The DHR system consists of 1) 3 dedicated DHR loops: the choice of 3 loops (3*100% redundancy) is made in assuming that one could be lost due to the accident initiating event (break for example) and that another one must be supposed unavailable (single failure criterion); 2) a metallic guard containment enclosing the primary system (referred as close containment), not pressurized in normal operation, having a free volume such as the fast primary helium expansion gives an equilibrium pressure of 1.0 MPa, in the first part of the transient (few hours). Each dedicated DHR loop designed to work in forced circulation with blowers or in natural circulation, is composed of 1) a primary loop (cross-duct connected to the core vessel), with a driving height of 10 meters between core and DHX mid-plan; 2) a secondary circuit filled with pressurized water at 1.0 MPa (driving height of 5 meters for natural circulation DHR); 3) a ternary pool, initially at 50 C. degrees, whose volume is determined to handle one day heat extraction (after this time delay, additional measures are foreseen to fill up the pool). The results obtained on the reliability of the DHR system and on the most important input parameters are very different from one scenario to the other showing the necessity for the PSA to perform specific reliability analysis of the passive system for each considered scenario. The analysis shows that the DHR system working in natural circulation is

  1. Flow inversion and natural convection in a MTR (Materials Testing Reactor)

    International Nuclear Information System (INIS)

    Gimenez, M.O.; Clausse, A.

    1990-01-01

    The thermohydraulic evolution of a refrigerating channel of the MTR (Materials Testing Reactors) RA-6 reactor's core, at the Bariloche Atomic Center, has been studied during the transient caused by the primary system's pump decommissioning. This transient constitutes one of the reactor's operating power boundaries due to the maximum temperature permissible in fuel plates. The problem regarding the thermohydraulic code altered for the rectangular geometry calculation characteristic of the MTR design is analyzed. (Author) [es

  2. Uncertainty correlation in stochastic safety analysis of natural circulation decay heat removal of liquid metal reactor

    International Nuclear Information System (INIS)

    Takata, Takashi; Yamaguchi, Akira

    2009-01-01

    Since various uncertainties of input variables are involved and nonlinearly-correlated in the Best Estimate (BE) plant dynamics code, it is of importance to evaluate the importance of input uncertainty to the computational results and to estimate the accuracy of the confidence level of the results. In order to estimate the importance and the accuracy, the authors have applied the stochastic safety analysis procedure using the Latin Hypercube sampling method to Liquid Metal Reactor (LMR) natural circulation Decay Heat Removal (DHR) phenomenon in the present paper. 17 input variables are chosen for the analyses and 5 influential variables, which affect the maximum coolant temperature at the core in a short period of time (several tens seconds), are selected to investigate the importance by comparing with the full-scope parametric analysis. As a result, it has been demonstrated that a comparative small number of samples is sufficient enough to estimate the dominant input variable and the confidence level. Furthermore, the influence of the sampling method on the accuracy of the upper tolerance limit (confidence level of 95%) has been examined based on the Wilks' formula. (author)

  3. Rapid solar-thermal dissociation of natural gas in an aerosol flow reactor

    International Nuclear Information System (INIS)

    Dahl, Jaimee K.; Buechler, Karen J.; Finley, Ryan; Stanislaus, Timothy; Weimer, Alan W.; Lewandowski, Allan; Bingham, Carl; Smeets, Alexander; Schneider, Adrian

    2004-01-01

    A solar-thermal aerosol flow reactor process is being developed to dissociate natural gas (NG) to hy drogen (H 2 ) and carbon black at high rates. Concentrated sunlight approaching 10 kW heats a 9.4 cm long x2.4 cm diameter graphite reaction tube to temperatures ∼2000 K using a 74% theoretically efficient secondary concentrator. Pure methane feed has been dissociated to 70% for residence times less than 0.1 s. The resulting carbon black is 20-40 nm in size, amorphous, and pure. A 5 million (M) kg/yr carbon black/1.67 M kg/yr H 2 plant is considered for process scale-up. The total permanent investment (TPI) of this plant is $12.7 M. A 15% IRR after tax is achieved when the carbon black is sold for $0.66/kg and the H 2 for $13.80/GJ. This plant could supply 0.06% of the world carbon black market. For this scenario, the solar-thermal process avoids 277 MJ fossil fuel and 13.9 kg-equivalent CO 2 /kg H 2 produced as compared to conventional steam-methane reforming and furnace black processing

  4. Study on applicability of PIV measurement to natural convection in a scaled reactor vessel model

    International Nuclear Information System (INIS)

    Murakami, Takahiro; Koga, Tomonari; Eguchi, Yuzuru; Watanabe, Osamu

    2009-01-01

    The applicability of Particle Image Velocimetry (PIV) to natural convection in the plenum of a scaled water test model of the Japan Sodium-cooled Fast Reactor (JSFR) is studied in the paper. PIV measurement of such a buoyancy-driven flow in a geometrically complicated vessel is difficult in general, because the detection rate of tracer particles tends to decrease, and the noisy optical reflection to increase. In our measurements, tracer particles are adequately seeded in the hot plenum and particle images are captured by using a double-pulsed Nd:YAG laser and a high-speed camera. Then, image-processing techniques are employed to eliminate unphysical velocity vectors and unnecessary background images. The PIV results have shown that clear flow pattern can be extracted by time-averaging 300 sets of instantaneous PIV data in spite of highly fluctuating features of velocity in space and time. Moreover, the evaluation of the statistical quantities such as variance, skewness, and kurtosis has revealed the characteristic of the non-stationary spouting flows at the heater outlet. (author)

  5. Neutronic design for a 100MW{sub th} Small modular natural circulation lead or lead-alloy cooled fast reactors core

    Energy Technology Data Exchange (ETDEWEB)

    Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q., E-mail: shchshch@ustc.edu.cn, E-mail: hlchen1@ustc.edu.cn, E-mail: kulah@mail.ustc.edu.cn, E-mail: zchen214@mail.ustc.edu.cn, E-mail: zengqin@ustc.edu.cn [Univ. of Science and Technology of China, School of Nuclear Science and Technology, Hefei, Anhui (China)

    2015-07-01

    Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW{sub th} natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

  6. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    International Nuclear Information System (INIS)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L.

    2015-01-01

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  7. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L., E-mail: rogerio.tdn@gmail.com, E-mail: souzalima_ca@ien.gov.br, E-mail: oliveira.afelipe@gmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: faccini@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  8. Study of natural circulation for the design of a research reactor using computational fluid dynamics and evolutionary computation techniques

    International Nuclear Information System (INIS)

    Oliveira, Andre Felipe da Silva de

    2012-01-01

    Safety is one of the most important and desirable characteristics in a nuclear plant Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid dynamics (CFD) code called CFX is used to simulate the process of natural circulation in a research reactor pool after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous medium, where the parameters were obtained from a separately detailed CFD analysis. This work also aims to study the viability of the implementation of Differential Evolution algorithm for optimization the physical and operational parameters that, obeying the laws of similarity, lead to a test section on a reduced scale of the reactor pool.

  9. Theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor

    International Nuclear Information System (INIS)

    Gou Junli; Qiu Suizheng; Su Guanghui; Jia Dounan

    2006-01-01

    This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation. (authors)

  10. From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca

    International Nuclear Information System (INIS)

    Stefanovic, D.; Pesic, M.

    1995-01-01

    A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

  11. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  12. The low-power low-pressure flow resonance in a natural circulation cooled boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, T.H.J.J. van der; Stekelenburg, A.J.C. [Delft Univ. of Technology (Netherlands)

    1995-09-01

    The last few years the possibility of flow resonances during the start-up phase of natural circulation cooled BWRs has been put forward by several authors. The present paper reports on actual oscillations observed at the Dodewaard reactor, the world`s only operating BWR cooled by natural circulation. In addition, results of a parameter study performed by means of a simple theoretical model are presented. The influence of relevant parameters on the resonance characteristics, being the decay ratio and the resonance frequency, is investigated and explained.

  13. Geochemical study of the insoluble organic material (kerogen) in the Oklo uranium ore and the associated Francevillian schists

    International Nuclear Information System (INIS)

    Vandenbroucke, M.; Rouzaud, J.N.; Oberlin, A.

    1978-01-01

    The purpose of this study was to describe the organic material associated with uranium ore and ore transformations undergone by it, in terms of the following problems: (1) In the natural reactor zones, evolution of the organic material in the core and as a function of the distance away from it; (2) Comparison of organic materials from a rich and a poor ore; (3) Intercomparison of organic materials in the dispersed and concentrated state; (4) Comparison of organic materials in the uranium ore zones and in the adjacent non-mineralized Francevillian. The organic material from the reactor core could not be isolated by the normal techniques of treatment with acid. It is found in other cases that the organic material is oxidized in the uranium-bearing sediments and that the nearer to the reaction zone, the greater the oxidation, irrespective of the state of dispersion of the organic material in the rock. The uranium content does not affect this phenomenon, which is attributed to the action of the water raised to a high temperature in the vicinity of the reaction zones. On the basis of the present distribution of organic material and uranium the authors suggest a pattern for the formation of the deposit that would take into account localization of the ore in the sandstones and the part played by organic material in the accumulation process. (author)

  14. Fundamental study on thermo-hydraulics during start-up in natural circulation boiling water reactors, (1)

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang Jing-Hsien; Takahashi, Tohru; Wataru, Masumi; Mori, Michitsugu.

    1992-01-01

    Recently, many concepts, in which passive and simplified functions are actively adapted, have been proposed for the next generation LWRs. The natural circulation BWR is one such considered from the requirements for next generation LWRs as compared with current BWRs. It is pointed out from this consideration that a thermo-hydraulic instability, which may appear during start-up, greatly influences concept feasibility because its occurence makes operation for raising power output difficult. Thermo-hydraulic instabilities are investigated experimentally under conditions simulating normal and abnormal start-up processes. It is clarified that three kinds of thermo-hydraulic instabilities may occur during start-up in the natural circulation BWR according to its procedure and reactor configuration, which are (1) geysering induced by condensation, (2) natural circulation instability induced by hydrostatic head fluctuation in steam separators and (3) density wave instability. Driving mechanisms of the geysering and the natural circulation instability, which have never understood enough, are inferred from the results. Finally, the difference of thermo-hydraulic behavior during start-up processes between thermal natural circulation boilers and the Dodewaard reactor is discussed. (author)

  15. Natural genetic transformation in Acinetobacter sp. BD413 Biofilms: introducing natural genetic transformation as a tool for bioenhancement of biofilm reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hendrickx, L

    2002-07-01

    This study focussed on the localization and quantification of natural genetic transformation using neutral and disadvantageous genes in monoculture biofilms to investigate gene transfer and expression of the transferred genes in the absence of a selective advantage. Data obtained by this investigation were regarded as initial steps for evaluating the applicability of adding catabolic traits into the indigenous bacterial community of biofilm reactors by in situ natural genetic transformation. Because Acinetobacter spp. strains are readily found in waste water treatment plants and because Acinetobacter sp. BD413 possesses a high effective level of competence, natural genetic transformation was investigated in monoculture Acinetobacter sp. BD413 biofilms. The genes used for transformation encoded for the green fluorescent protein (GFP) and its variants. Monitoring of transformation events were performed with the use of automated confocal laser scanning microscopy (CLSM) and semi automated digital image processing and analysis. (orig.)

  16. Thermo-mechanical behaviour of FBTR reactor vessel due to natural convection in cover gas space

    International Nuclear Information System (INIS)

    Srinivasan, G.; Varadarajan, S.; Kapoor, R.P.

    1988-01-01

    Fast Breeder Test Reactor is a 40 MW(t), loop type sodium cooled reactor, similar in design to Rapsodie. The Reactor Assembly, which is the heart of FBTR, comprises the Reactor Vessel (RV) housed in a safety vessel within a concrete cell (A1 Cell). The RV which supports the core is shielded at the top by two rotatable plugs which are stacked with layers of borated graphite and steel. The smaller plug (SRP), is mounted excentric to the larger one (LRP). A nominal annular gap of 16 mm is provided between RV and LRP and between LRP and SRP to enable free rotation of the plugs. Stainless Steel insulation is fixed inside the steel vessel, to avoid overheating of the A1 Cell concrete. The core is supported by the Grid Plate (GP), bolted to the RV. During preheating, sodium charging and isothermal runs upto 350 0 C, temperature asymmetries were noticed in the reactor vessel wall in the cover gas space. This was attributable to convection currents in the annulus between RV and LRP. The asymmetries also resulted in a lateral shift of the grid plate. This paper discusses our experience in suppressing these convection currents, and minimising the grid plate shift

  17. Evolution of on-power fuelling machines on Canadian natural uranium power reactors

    International Nuclear Information System (INIS)

    Isaac, P.

    1984-10-01

    The evolution of the on-power fuel changing process and fuelling machines on CANDU heavy-water pressure tube power reactors from the first nuclear power demonstration plant, 22 MWe NPD, to the latest plants now in design and development is described. The high availability of CANDU's is largely dependent on on-power fuelling. The on-power fuelling performance record of the 16 operating CANDU reactors, covering a 22 year period since the first plant became operational, is given. This shows that on-power fuel changing with light (unshielded), highly mobile and readily maintainable fuelling machines has been a success. The fuelling machines have contributed very little to the incapabilities of the plants and have been a key factor in placing CANDUs in the top ten list of world performance. Although fuel handling technology has reached a degree of maturity, refinements are continuing. A new single-ended fuel changing concept for horizontal reactors under development is described. This has the potential for reducing capital and operating costs for small reactors and increasing the fuelling capability of possible large reactors of the future

  18. New studies of the natural convection around a fuel rod of the BME training reactor with PIV/LIF technique

    International Nuclear Information System (INIS)

    Szijarto, R.; Aszodi, A.; Yamaji, B.

    2011-01-01

    In this paper the model of a fuel pin of the Training Reactor of Budapest University of Technology and Economics was investigated with Particle Image Velocimetry and Laser Induced Fluorescence measurement methods. An experimental setup was designed, built and optimized to investigate the natural convection around a model of a fuel pin of the Training Reactor. The processes were analysed using an electrically heated rod, which models the geometry of the fuel rods in the Training Reactor. The heated length of the model is the same as the active length of the real fuel rods. The rod is placed in a glass tank with a shape of a square-based prism. An additional cooling system ensures constant flow conditions around the rod. The setup consists of an additional flow channel box, the equivalent diameter of which is equal to the equivalent diameter of the real fuel assembly. Simultaneous measurements of velocity and temperature fields were performed in different vertical positions for both cases of natural convection with and without the flow channel box. The effect of the presence of the channel was analyzed, and a laminarizating influence was observed. The local heat transfer coefficient was calculated for every measurement. The two dimensional measurement techniques gave extensive results, the structure of the hydraulic and thermal boundary layer were fully analyzed. (Authors)

  19. Solutions for Foaming Problems in Biogas Reactors Using Natural Oils or Fatty Acids as Defoamers

    DEFF Research Database (Denmark)

    Kougias, Panagiotis; Boe, Kanokwan; Angelidaki, Irini

    2015-01-01

    Foaming is one of the most common and important problems in biogas plants, leading to severe operational, economical, and environmental drawbacks. Because addition of easily degradable co-substrates for boosting the biogas production can suddenly raise the foaming problem, the full-scale biogas...... results from our previous extensive research along with some unpublished data on defoaming by rapeseed oil and oleic acid in manure-based biogas reactors. It was found that both compounds exhibited remarkable defoaming efficiency ranging from 30 to 57% in biogas reactors suffering from foaming problems...... promoted by the addition of protein, lipid, or carbohydrate co-substrates. However, in most cases, the defoaming efficiency of rapeseed oil was greater than that of oleic acid, and therefore, rapeseed oil is recommended to be used in biogas reactors to solve foaming problems....

  20. Investigation of natural convection in Miniature Neutron Source Reactor of Isfahan by applying the porous media approach

    Energy Technology Data Exchange (ETDEWEB)

    Abbassi, Yasser, E-mail: y.abbassi@mihanmail.ir [Department of Engineering, University of Shahid Beheshti, Tehran (Iran, Islamic Republic of); Asgarian, Shahla [Department of Chemical Engineering, Isfahan University, Tehran (Iran, Islamic Republic of); Ghahremani, Esmaeel; Abbasi, Mohammad [Department of Engineering, University of Shahid Beheshti, Tehran (Iran, Islamic Republic of)

    2016-12-01

    Highlights: • We carried out a CFD study to investigate transient natural convection in MNSR. • We applied porous media approach to simplify the complex core of MNSR. • Method have been verified with experimental data. • Temperature difference between the core inlet and outlet has been obtained. • Flow pattern and temperature distribution have been presented. - Abstract: The small and complex core of Isfahan Miniature Neutron Source Reactor (MNSR) in addition to its large tank makes a parametric study of natural convection difficult to perform in aspects of time and computational resources. In this study, in order to overcome this obstacle the porous media approximation has been used. This numerical technique includes two steps, (a) calculation of porous media variables such as porosity and pressure drops in the core region, (b) simulation of natural convection in the reactor tank by assuming the core region as a porous medium. Simulation has been carried out with ANSYS FLUENT® Academic Research, Release 16.2. The core porous medium resistance factors have been estimated to be, D{sub ij} = 1850 [1/m] and C{sub ij} = 415 [1/m{sup 2}]. Natural Convection simulation with Boussinesq approximation and variable property assumption have been performed. The experimental data and nuclear codes available in the literature, have verified the method. The average temperature difference between the experimental data and this study results was less than 0.5 °C and 2.0 °C for property variable technique and Boussinesq approximation, respectively. Temperature distribution and flow pattern in the entire reactor have been obtained. Results have shown that the temperature difference between core outlet and inlet is about 18°C and in this situation flow rate is about 0.004 kg/s. A full parametric study could be the topic of future investigations.

  1. Radio-active pollution near natural uranium-graphite-gas reactors

    International Nuclear Information System (INIS)

    Chassany, J.; Pouthier, J.; Delmar, J.

    1967-01-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [fr

  2. Post shut-down decay heat removal from nuclear reactor core by natural convection loops in sodium pool

    Energy Technology Data Exchange (ETDEWEB)

    Rajamani, A. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Sundararajan, T., E-mail: tsundar@iitm.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Prasad, B.V.S.S.S. [Department of Mechanical Engineering, Indian Institute of Technology Madras, Chennai 600036 (India); Parthasarathy, U.; Velusamy, K. [Nuclear Engineering Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India)

    2016-05-15

    Highlights: • Transient simulations are performed for a worst case scenario of station black-out. • Inter-wrapper flow between various sub-assemblies reduces peak core temperature. • Various natural convection paths limits fuel clad temperatures below critical level. - Abstract: The 500 MWe Indian pool type Prototype Fast Breeder Reactor (PFBR) has a passive core cooling system, known as the Safety Grade Decay Heat Removal System (SGDHRS) which aids to remove decay heat after shut down phase. Immediately after reactor shut down the fission products in the core continue to generate heat due to beta decay which exponentially decreases with time. In the event of a complete station blackout, the coolant pump system may not be available and the safety grade decay heat removal system transports the decay heat from the core and dissipates it safely to the atmosphere. Apart from SGDHRS, various natural convection loops in the sodium pool carry the heat away from the core and deposit it temporarily in the sodium pool. The buoyancy driven flow through the small inter-wrapper gaps (known as inter-wrapper flow) between fuel subassemblies plays an important role in carrying the decay heat from the sub-assemblies to the hot sodium pool, immediately after reactor shut down. This paper presents the transient prediction of flow and temperature evolution in the reactor subassemblies and the sodium pool, coupled with the safety grade decay heat removal system. It is shown that with a properly sized decay heat exchanger based on liquid sodium and air chimney stacks, the post shutdown decay heat can be safely dissipated to atmospheric air passively.

  3. A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

    International Nuclear Information System (INIS)

    Reed, Mark; Parker, Ronald R.; Forget, Benoit

    2012-01-01

    This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more

  4. Nature and characteristics of pulsing flow in trickle-bed reactors

    NARCIS (Netherlands)

    Boelhouwer, J.G.; Piepers, H.W.; Drinkenburg, A.A.H.

    2002-01-01

    Pulsing flow is well known for its advantages in terms of an increase in mass and heat transfer rates, complete catalyst wetting and a decrease in axial dispersion compared to trickle flow. The operation of a trickle-bed reactor in the pulsing flow regime is favorable in terms of a capacity increase

  5. Solar membrane natural gas steam-reforming process: evaluation of reactor performance

    NARCIS (Netherlands)

    de Falco, M.; Basile, A.; Gallucci, F.

    2010-01-01

    In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

  6. Solar membrane natural gas steam-reforming process : evaluation of reactor performance

    NARCIS (Netherlands)

    Falco, de M.; Basile, A.; Gallucci, F.

    2010-01-01

    In this work, the performance of an innovative plant for efficient hydrogen production using solar energy for the process heat duty requirements has been evaluated via a detailed 2D model. The steam-reforming reactor consists of a bundle of coaxial double tubes assembled in a shell. The annular

  7. Conceptual analysis of a preliminary model for instability study in normal operation of a natural circulation reactor type EBWR, using Relap5/Mod 3.2

    International Nuclear Information System (INIS)

    Ojeda S, J.; Morales S, J.; Chavez M, C.

    2009-10-01

    This work intends a model using the code Relap5/Mod 3.2, for the instability study in normal operation of a natural circulation reactor type ESBWR. A conceptual analysis is considered because all the information was obtained of the open literature, and some of reactor operation or dimension (not available) parameters were approached. As starting point was took the pattern developed for reactor type BWR, denominated Browns Ferry and changes were focused in elimination of bonds of forced recirculation, in modification of operation parameters, dimensions and own control parameters, according to internal code structure. Additionally the nodalization outline is described analyzing for separate the four fundamental areas employees in peculiar geometry of natural circulation reactor. Comparative analysis of results of stability behavior obtained with those reported in the open literature were made, by part of commercial reactor designer ESBWR. (Author)

  8. Theoretical and experimental research of natural convection in the core of the high temperature pebble bed reactor

    International Nuclear Information System (INIS)

    Schuerenkraemer, M.

    1984-04-01

    The physical model of the developed THERMIX-2D-code for computing thermohydraulic behaviour of the core of high temperature pebble bed reactors is verified by experiments with natural convection flow. Such fluid flow behaviour can be of very high importance for the real reactor in the case of natural heat removal decay. The experiments are performed in a special set up testing-stand with pressures up to 30 bars and temperatures up to 300 0 C by using air and helium as fluid. In comparison with the experimental data the numerical results show that a good and useful simulation is given by the program. Pure natural convection flow in packed pebble beds is calculated with a very high degree of reliability. The investigation of flow stability demonstrate that radial-symmetric relations are not given temporarily when national convection is overlayed by forced convection flow. In the discussion it is explained when and to what extent the program leds to useful results in such situations. The test of the effective heat conductivity lambdasub(eff) results in an improvement of the lambdasub(eff)-data used so far for temperatures below 1300 0 C. (orig.) [de

  9. Numerical research on natural convection in molten salt reactor with non-uniformly distributed volumetric heat generation

    International Nuclear Information System (INIS)

    Qian Libo; Qiu Suizheng; Zhang Dalin; Su Guanghui; Tian Wenxi

    2010-01-01

    Molten salt reactor is one of the six Generation IV systems capable of breeding and transmutation of actinides and long-lived fission products, which uses the liquid molten salt as the fuel solvent, coolant and heat generation simultaneously. The present work presents a numerical investigation on natural convection with non-uniform heat generation through which the heat generated by the fluid fuel is removed out of the core region when the reactor is under post-accident condition or zero-power condition. The two-group neutron diffusion equation is applied to calculated neutron flux distribution, which leads to non-uniform heat generation. The SIMPLER algorithm is used to calculate natural convective heat transfer rate with isothermal or adiabatic rigid walls. These two models are coupled through the temperature field and heat sources. The peculiarities of natural convection with non-uniform heat generation are investigated in a range of Ra numbers (10 3 ∼ 10 7 ) for the laminar regime of fluid motion. In addition, the numerical results are also compared with those containing uniform heat generation.

  10. Preliminary definition of the design of a nuclear reactor for research and radioisotope production using natural uranium and heavy water

    International Nuclear Information System (INIS)

    Llagostera Beltran, J.I.

    1982-01-01

    A study was conducted about the evolution of the Brazilian importations of radioisotopes, from the beginning of the 70's since they have been increasingly used in the Country. In view of the limited production capacity of radioactive isotopes now existing in Brazil, a nuclear reactor type (natural uranium and heavy water) was defined, for research and production of radioisotopes, wich, besides providing, at least partially, the Brazilian needs of said isotopes, permits a large national participation in its project, construction and operating maintenance. The processes for heavy water production have been analyzed and it could be detected what is the best alternative for the production thereof, in low scale, in Brazil. The options concerning the definition of the main components of the reactor were justified and its most important features were determined, in relation to the neutronic and thermal aspects, being so defined its most significant parameters. The annual quantities were estimated, in terms of total and specific activity, for the radioisotopes that could be obtained by means of the proposed reactor, which, by now, are participating, to a large extent, in the total of Brazilian importation of radioactive isotopes. (Author) [pt

  11. To the safety conception of the high temperature reactor with natural heat removal decay in teh case of accidents

    International Nuclear Information System (INIS)

    Petersen, K.

    1983-10-01

    On September 22, 1970, for the first time an accident simulation experiment with complete failure of the forced core cooling and the nuclear shut-down system was performed in the AVR-reactor: Due to a small heat-up of the fuel the nuclear chain-reaction was interrupted and an overheating of the core and structure was prevented due to the natural heat-convection. On the basis of the meanwhile developed computer-methods and accompanying experimental investigations it is now possible to determine exactly the behaviour of the non actively controlled core of the high temperature reactor, and to understand better the course of the AVR-experiments. On the same basis the potential and the limits of the safety conception realized in the AVR with self-stabilization in the case of accident can be determined. Such a small high temperature reactor as for example the HTR-modul of the KWU, which is characterized by a reliable and simple safety-technique with a minimum of expensive active systems, can be realized using a 2-zone-core up to a unit size of nearly 250 MW(th). (orig.) [de

  12. Natural repository analouge program. Progress report, January 1-March 31, 1981

    International Nuclear Information System (INIS)

    Curtis, D.B.

    1981-05-01

    Samples from Oklo Reactor zone-9 (ORZ-9) have been analyzed for the isotopic abundances of Nd, Ce, Ru, and Mo. Interpretation of the Nd data has begun as part of the effort to reconstruct the operating parameters of the reactor. The study of ORZ-9 and the peripheral rocks is being enhanced by additional analytical capabilities. A procedure was developed to measure uranium isotopic ratios with high precision. This new method was used for the analysis of rocks peripheral to ORZ-9. Two rocks containing relatively small quantities of uranium were depleted in 235 U. The result demonstrates that small quantities of uranium were removed from the reactor zone and redistributed over distances of several tens of meters. Procedures are being designed to make high precision measurements of the relative abundances of barium isotopes. They will be used as part of a study of the transport of alkali and alkaline earth elements at Oklo. Samples from distances up to 300 meters from the known mineralized area at Oklo have been selected and prepared in an effort to identify element transport paths over longer distances. A sample from the Athabasca sandstone, overlying the uranium ores at Key Lake, and another from the transition zone at the unconfromity between the sandstone and the basement were subjected to sequential leaches designed to preferentially dissolve specific minerals. Lead isotopic analyses on the leachs yielded two sets of data, which indicate the loss of uranium or the addition of lead from a radiogenic source early in the geologic history of the rocks

  13. Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Takaaki; Enuma, Yasuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwasaki, Takashi [Nuclear Energy System Inc., Tokyo (Japan); Ohyama, Kazuhiro [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2001-05-01

    The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

  14. Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

    International Nuclear Information System (INIS)

    Choi, A-Reum; Song, Hyuk-Jin; Park, Jong-Woon

    2015-01-01

    During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

  15. Thermal hydraulic aspects of steam drum level control philosophy for the natural circulation based heavy water reactor

    International Nuclear Information System (INIS)

    Gupta, S.K.; Gaikwad, A.J.; Kumar, Rajesh

    2004-01-01

    From safety considerations advanced nuclear reactors rely more and more on passive systems such as natural circulation for primary heat removal. A natural circulation based water reactor is relatively larger in size so as to reduce flow losses and channel type for proper flow distribution. From the size of steam drum considerations it has to be multi loop but has a common inlet header. Normally the turbine follows the reactor. This paper addresses the thermal hydraulic aspects of the steam drum pressure and level control philosophy for a four drum, natural circulation based, channel type boiling water advanced reactor. Three philosophies may be followed for drum control viz. individual drum control, one control drum approach and an average of all the four drums. For drum pressure control, the steam flow to the turbine is be regulated. A single point pressure control is better than individual drum pressure control. This is discussed in the paper. But the control point has to be at a place down steam the point where all steam line from individual drum meet. This may lead to different pressure in all the four drums depending on the power produced in the respective loops. The difference in pressure cannot be removed even if the four drums are directly connected through pipes. Also the pressure control scheme with/without interconnection is discussed. For level, the control of individual drum may not be normally possible because of common inlet header. As the frictional pressure drops in the large diameter downcomers are small as compared to elevation pressure drops, the level in all the steam drum tend to equalize. Consequently a single representative drum level may be chosen as a control variable for controlling level in all the four drums. But in case, where all the four loops are producing different powers and single point pressure control is effective, the scheme may not work satisfactorily. the level in a drum may depend on the power produced in the loop

  16. A study of thermal-hydraulic requirements for increasing the power rates for natural-circulation boiling water reactors

    International Nuclear Information System (INIS)

    Yasuo, A.; Inada, F.; Hidaka, M.

    1992-01-01

    In this paper, the feasibility of higher power rates for natural-circulation boiling water reactors (BWRs) is studied with the objective of examining the flexibility of the plant power rate in constructing such plants to cope with the increasing demand for electricity. By applying existing one-dimensional design codes, the riser heights necessary to meet two major thermal-hydraulic requirements, i.e., critical power and core stability, are systematically calculated. Several restrictions on the maximum diameter and height of the pressure vessel are also considered because these restrictions could make construction impossible or drastically increase the construction costs. A very simple map of the dominant parameters for higher power rates is obtained. It is concluded that natural-circulation BWRs of >1000 MW (electric) will be feasible within the restrictions considered here

  17. Evaluations of two-phase natural circulation flow induced in the reactor vessel annular gap under ERVC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon, E-mail: tomo@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Cheung, Fan-Bill [The Pennsylvania State University, University Park, PA 16802 (United States); Park, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two-phase natural circulation flow induced in insulation gap was investigated. Black-Right-Pointing-Pointer Half-scaled non-heating experiments were performed to evaluate flow behavior. Black-Right-Pointing-Pointer The loop-integrated momentum equation was formulated and solved asymptotically. Black-Right-Pointing-Pointer First-order approximate solution was obtained and agreed with experimental data. - Abstract: The process of two-phase natural circulation flow induced in the annular gap between the reactor vessel and the insulation under external reactor vessel cooling conditions was investigated experimentally and analytically in this study. HERMES-HALF experiments were performed to observe and quantify the induced two-phase natural circulation flow in the annular gap. A half-scaled non-heating experimental facility was designed by utilizing the results of a scaling analysis to simulate the APR1400 reactor and its insulation system. The behavior of the boiling-induced two-phase natural circulation flow in the annular gap was observed, and the liquid mass flow rates driven by the natural circulation loop and the void fraction distribution were measured. Direct flow visualization revealed that choking would occur under certain flow conditions in the minimum gap region near the shear keys. Specifically, large recirculation flows were observed in the minimum gap region for large air injection rates and small outlet areas. Under such conditions, the injected air could not pass through the minimum gap region, resulting in the occurrence of choking near the minimum gap with a periodical air back flow being generated. Therefore, a design modification of the minimum gap region needs to be done to facilitate steam venting and to prevent choking from occurring. To complement the HERMES-HALF experimental effort, an analytical study of the dependence of the induced natural circulation mass flow rate on the inlet area and the

  18. Comparative study on aerosol removal by natural processes in containment in severe accident for AP1000 reactor

    International Nuclear Information System (INIS)

    Sun, Xiaohui; Cao, Xinrong; Shi, Xingwei; Yan, Jin

    2017-01-01

    Highlights: • Characteristics of aerosol distribution in containment are obtained. • Aerosol removal by natural processes is comparative studied by two methods. • Traditional rapid assessment method is conservative and can be applied in AP1000 reactor. - Abstract: Focusing on aerosol removal by naturally occurring processes in containment in severe accident for AP1000, integral severe accident code MELCOR and rapid assessment method mentioned in NUREG/CR-6189 are utilized to study aerosol removal by natural processes, respectively. Three typical severe accidents, induced by large break loss of coolant accident (LBLOCA), small break loss of coolant accident (SBLOCA) and steam generator tube rupture (SGTR), respectively, are selected for the study. The results obtained by two methods were further compared in the following several aspects: efficiency of aerosol removal by natural processes, peak time of aerosol suspended in containment atmosphere, peak amount of aerosol suspended in containment atmosphere, time when aerosol removal efficiency by natural processes is up to 99.9%. It was further concluded that results obtained by rapid assessment with shorter calculation process are more conservative. The analysis results provide reference to assessment method selection of severe accident source term for AP1000 nuclear emergency.

  19. Tests of the heat transfer characteristic of air cooler during cooling by natural convection of the Fast Breeder Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purpose of this study is to confirm the heat transfer characteristics of the air cooler (AC) of the Fast Breeder Reactor(FBR) which has a function to remove the residual heat of the reactor by heat exchange between sodium and air in natural convection region if electric power would be lost. In order to confirm the characteristics of the AC installed in the FBR plant, the heat transfer test by using the AC which is installed in the sodium test loop owned by Toshiba Corporation has been planned. In this study, the heat transfer characteristic tests were performed by using the AC in sodium test loop, and the CFD analyses were conducted to evaluate the test results and the heat transfer characteristics of the plant scale AC at the condition of natural convection. In addition, the elemental tests to confirm the influence of the heat transfer tube placement by using the heat transfer tube of the same specification as the AC of Monju were performed. (author)

  20. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  1. Influence of operation of national experimental nuclear reactor on the natural environment

    Directory of Open Access Journals (Sweden)

    Agnieszka Kaczmarek-Kacprzak

    2012-09-01

    Full Text Available This paper presents the impact of experimental nuclear reactor operations on the national environment, based on assessment reports of the radiological protection of active nuclear technology sources. Using the analysis of measurements carried out in the last 15 years, the trends are presented in selected elements of the environment on the Świerk Nuclear Centre site and its surroundings. In addition, the impact of research results is presented from the fi fteen year period of environmental analysis on building public confi dence on the eve of the start of construction of the first Polish nuclear power plant.

  2. The experimental study on the mass transfer model of boron injection for natural circular heating reactor

    International Nuclear Information System (INIS)

    Zha Meisheng; Nie Mengchen; Zhou Huizhong; Wang Liqun; Guo Weiping; Liu ZHiyong

    1989-09-01

    A pulse injection stimulus-response technique to study the boron mixing and transport performance after boron-loaded liquid was injected into the reactor core is described. The experiment was carried out in a simulation device. The simulation medium was used. The experimental results show that the lower plenum where the injection point located can be simplified to one scale inertial unit and the movement of boron mixture was only transported after it had entered into the fuel elements. The definition of boron initiative mixing fraction η is also given. By using relating data a dimensionless equation is obtained

  3. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    International Nuclear Information System (INIS)

    Monado, F.; Permana, S.

    2013-01-01

    Full-text: A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8 % HM. From the neutronic point of view, this design is in compliance with good performance. (author)

  4. Conceptual design study on very small long-life gas cooled fast reactor using metallic natural Uranium-Zr as fuel cycle input

    International Nuclear Information System (INIS)

    Monado, Fiber; Ariani, Menik; Su'ud, Zaki; Waris, Abdul; Basar, Khairul; Permana, Sidik; Aziz, Ferhat; Sekimoto, Hiroshi

    2014-01-01

    A conceptual design study of very small 350 MWth Gas-cooled Fast Reactors with Helium coolant has been performed. In this study Modified CANDLE burn-up scheme was implemented to create small and long life fast reactors with natural Uranium as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. The core with metallic fuel based was subdivided into 10 regions with the same volume. The fresh Natural Uranium is initially put in region-1, after one cycle of 10 years of burn-up it is shifted to region-2 and the each region-1 is filled by fresh Natural Uranium fuel. This concept is basically applied to all axial regions. The reactor discharge burn-up is 31.8% HM. From the neutronic point of view, this design is in compliance with good performance

  5. Characterization of natural circulation looping of emergency cooling systems in naval and advanced reactors

    International Nuclear Information System (INIS)

    Macedo, Luiz Alberto; Baptista Filho, Benedito Dias

    2000-01-01

    This paper describes the natural circuit looping, resumes the main project characteristics, presents results of the hydraulic characterization, consisting of pressure loss measurements, and presents results from calibration tests of the power and flow measurements and the first experiments in natural circulation. Those experiments comprised transients in natural circulation with application of application of power steps. The results shown a non linear behaviour of the magnetic flow meter and a dependence on the fluid temperature as well. The assembly circuit/instrumentation/data acquisition system is suitable for the research on emergency cooling passive systems

  6. A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

    International Nuclear Information System (INIS)

    Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

    1995-01-01

    The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

  7. Validation of RELAP5 model of experimental test rig simulating the natural convection in MTR research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khedr, A.; Abdel-Latif, Salwa H. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt); Abdel-Hadi, Eed A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; D' Auria, F. [Pisa Univ. (Italy)

    2016-03-15

    In an attempt to understand the built-up of natural circulation in MTR pool type upward flow research reactors after loss of power, an experimental test rig was built to simulate the loop of natural circulation in MTR reactors. The test rig consisting of two vertically oriented branches, in one of them the core is simulated by two rectangular, electrically heated, parallel channels. The other branch simulates the part of the return pipe that participates in the development of core natural circulation. In the first phase of the work, many experimental runs at different conditions of channel's power and branch's initial temperatures are performed. The channel's coolant and surface temperatures were measured. The measurements and their interpretation were published by the first three authors. In the present work the thermal hydraulic behavior of the test rig is complemented by theoretical analysis using RELAP5 Mod 3.3 system code. The analysis consisting of two parts; in the first part RELAP5 model is validated against the measured values and in the second part some of the other not measured hydraulic parameters are predicted and analyzed. The test rig is typically nodalized and an input dick is prepared. In spite of the low pressure of the test rig, the results show that RELAP5 qualitatively predicts the thermal hydraulic behaviour and the accompanied phenomenon of flow inversion of such facilities. Quantitatively, there is a difference between the predicted and measured values especially the channel's surface temperature. This difference may be return to the uncertainties in initial conditions of experimental runs, the position of the thermocouples which buried inside the heat structure, and the heat transfer package in RELAP5.

  8. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

    International Nuclear Information System (INIS)

    Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  9. NATBWR: a steady-state model for natural circulation in boiling-water reactors

    International Nuclear Information System (INIS)

    Healzer, J.M.; Abdollahian, D.

    1983-02-01

    This report documents the NATBWR steady-state BWR natural-circulation model and activities under EPRI Project RP1561-1 to gather data and predict the natural-circulation operation of the BWR. The report is organized into two parts, with the first part describing the NATBWR model and applications of the model to available BWR natural-circulation data and the second part providing user and programming information about the model. Five different operating BWR's were selected to demonstrate the application of the NATBWR model, one of each type from BWR/1 through BWR/4. For each operating plant, the available natural circulation data has been compared to model predictions. In addition to the data predictions, the behavior of the BWR system at reduced inventory, where the system is isolated and scrammed, and cooling provided by natural circulation has been analyzed. Finally, included as an appendix to Part 1 of this report is a discussion of the stability of the BWR system at natural-circulation conditions

  10. Effect of turbulent natural convection on sodium pool combustion in the steam generator building of a fast breeder reactor

    International Nuclear Information System (INIS)

    Karthikeyan, S.; Sundararajan, T.; Shet, U.S.P.; Selvaraj, P.

    2009-01-01

    A computational model is proposed to simulate sodium pool combustion considering the effect of turbulent natural convection in a vented enclosure of the steam generator building (SGB) of a fast breeder reactor. The model is validated by comparing the simulated results with the experimental results available in literature for sodium pool combustion in a CSTF vessel. After validation, the effects of vents and the location of the pool on the burning rate of sodium and the associated heat transfer to the walls are studied in an enclosure comparable in size to one floor of the steam generator building. In the presence of ventilation, the burning rate of sodium increases, but the total heat transferred to the walls of the enclosure is reduced. It is also found that the burning rate of sodium pool and the heat transfer to the walls of the enclosures vary significantly with the location of sodium pool.

  11. Influence of natural convection and diluent inerting on H2 and CO oxidation in the reactor cavity

    International Nuclear Information System (INIS)

    Wong, C.C.

    1988-01-01

    The question of complete in-cavity oxidation of combustible gases produced by core-concrete interactions following vessel breach has been investigated. It is overly optimistic to assume a complete oxidation because a variety of phenomena, such as steam inerting and oxygen transport by natural convection, may influence the degree of in-cavity oxidation that takes place. HECTR analyses of an ice-condenser containment during an S2HF drain-closed accident show that the in-cavity oxidation process is limited by the rate at which oxygen is transported into the reactor cavity region. Accumulation and subsequent combustion of hydrogen and carbon monoxide in the upper and lower compartments generate a peak pressure of 384 kPa (56 psig) at 7.4 h, that an earlier IDCOR analysis did not predict. (orig.)

  12. A comparative design study of PB-BI cooled reactor cores with forced and natural convection cooling

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Enuma, Yasuhiro; Tanji, Mikio

    2003-01-01

    A comparative core design study is performed on Pb-Bi cooled reactors with forced and natural convection (FC and NC) cooling. Major interests of the study are core performance and core safety features. The designed core concepts with nitride fuel achieve reasonable breeding capability. The results of unprotected event analyses such as UTOP and ULOF show that both of concepts have possible features to withstand unprotected events due to negative reactivity feedback by Doppler effect, control rod drive line expansion, etc. These results lead to a conclusion that both of concepts have possible capability as one of future promising core concepts. A FC cooling core concept has more advantage if fuel recycle viewpoint is emphasized. (author)

  13. Optimisation of the flow path in a conceptual pool type reactor under natural circulation with lead coolant

    International Nuclear Information System (INIS)

    Thiele, R.; Anglart, H.

    2014-01-01

    This contribution investigates the effects of a bypass flow blocking bottom plate and the influence of the heat transfer between the hot and cold leg in a small pool type reactor cooled through natural convection with lead coolant. The computations are carried out using 3D computational fluid dynamics, where small-detail parts, such as the core and heat exchangers are modeled using a porous media approach. The introduction of full conjugate heat transfer shows that the heat transfer between the hot and cold leg can deteriorate flow in the cold leg and lead to recirculation zones. These zones become even more pronounced with the introduction of a bottom plate, which on the other hand also increases the flow through the core and lowers the maximum temperature in the core by approximately 150 K. Based on the results, redesign suggestions for the bottom plate and the internal wall are made. (author)

  14. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2000-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. For example, the high cycle efficiency can be expected because of the similarity of the present cycle to the Ericsson cycle. Sodium-Water Interaction problem can be excluded by proper combination of the working fluids. As the economical feature, the present system is so simple that the liquid-metal main circular pump, the steam turbine generator, and even the steam generator can be excluded if the thermodynamic working fluid is injected directly into the high temperature liquid metal MHD working fluid. In addition, the present system has the potential to be applied to various heat sources including solar energy because of the high flexibility of the operation temperature. In the present paper, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It is found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It is, however, found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. As the conclusions, it is recommended to perform experimental study to obtain the fundamental data, such as the gas-liquid slip ratio in the high-density liquid-metal two-phase natural circulation. (author)

  15. Reduced scaling of thermal-hydraulic circuits for studies of PWR reactors natural circulation

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1993-01-01

    The Ishii et al. hydrodynamic similarity criteria for natural circulation were used for scaling reduced models of prototype passive residual heat removal system of a 600 M We PWR. The physical scales of the thermohydraulic parameters obtained presented a reasonable agreement when compared with simplified analytic models of the systems. (author)

  16. Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. Final Report of a Coordinated Research Project 2008-2012

    International Nuclear Information System (INIS)

    2014-11-01

    The IAEA supports Member States in the area of advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA support, and ensures that all technical activities are in line with expressed needs of Member States. Among this broad range, the IAEA proposes and establishes coordinated research projects (CRPs), aimed at improving Member State capability in fast reactor design and analysis. An important opportunity to perform collaborative research activities was provided by the system startup tests carried out by the Japan Atomic Energy Agency (JAEA) in the prototype loop type sodium cooled fast reactor Monju, in particular a turbine trip test performed in December 1995. As the JAEA opened the experimental dataset to international collaboration in 2008, the IAEA launched the CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. The CRP, together with eight institutes from seven States, has contributed to improving capabilities in sodium cooled fast reactors simulation through code verification and validation, with particular emphasis on thermal stratification and natural circulation phenomena

  17. Natural convection cooling of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Akhtar, K.M.

    1991-08-01

    The first high power and equilibrium LEU cores of PARR-1 have been analysed to assess the maximum operating power based on natural convection cooling, need for forced cooling to remove the decay heat and to estimate safety margins that commensurate with the predetermined power limit. Computer code NATCON and standard correlations have been used for the analysis. The parameters studied includes coolant velocity, temperature distribution in the core, heat fluxes at onset of nucleate boiling, pulsed boiling and burnup. (author)

  18. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  19. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    International Nuclear Information System (INIS)

    Massoud, M.

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients

  20. An experimental study on the two-phase natural circulation flow through the gap between reactor vessel and insulation under ERVC

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang-Soon; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik; Kim, Hwan-Yeol; Kim, Hee-dong

    2005-04-01

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, T-HERMES-SMALL and HERMES-HALF experiments have been performed. For the T-HERMES-SMALL experiments, an 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper outlet area and configuration, and water head condition. The experimental data were also compared with numerical ones given by simple loop analysis. And non-heating small-scaled experiments have also been performed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition. The HERMES-HALF experiment is a half-scaled / non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The behaviors of the two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the air injection rate, the coolant inlet area and configuration, and the outlet area and also the water head condition of coolant reservoir. From the experimental flow observation, the recirculation flows in the near region of the shear key were identified. At a higher air injection rate condition, higher recirculation flows and choking phenomenon in the near region of the shear key were observed. As the water inlet areas increased, the natural circulation mass flow rates asymptotically increased, that is, they converged at a specific value. And the experimental correlations for the natural circulation mass flow rates along with a variation of the inlet / outlet area and wall heat flux were

  1. Artificial and natural radioactivity measurements in the vicinity of Ghana nuclear research reactor (GHARR-1)

    International Nuclear Information System (INIS)

    Faanu, A.; Awudua, A.R.; Darko, E.O.; Emi-Reynolds, G.; Inkooma, S.; Adukpo, O.; Kpeglo, D.O.; Lawluva, H.; Obeng, M.K; Titiati, J.; Agyeman, B.; Kpodzro, R.; Ibrahim, A.; Gloverb, E.T.

    2010-01-01

    Radioactivity concentrations of 226Ra, 232Th, 40K and 137Cs in soil and water samples around the Ghana Research Reactor-1 (GHARR-1) and the immediate surroundings have been investigated using gamma spectrometry. The primary aim of this study was to establish baseline radioactivity levels in the environs of GHARR-1. The average activity concentration in soil for 226Ra, 232Th, 40K and 137Cs were 19.8 Bqkg-1, 40.4 Bqkg-1, 95.3 Bqkg-1 and 1.5 Bqkg-1 respectively. For the water samples the average activity concentration of 226Ra was 2.15 Bql-1, 232Th was 0.61 Bql-1, 40K was 10.75Bql-1 and 137Cs was 0.47 Bql-1. The 226Ra and 232Th concentrations compare quite well with world averages, whilst the 40K concentration was lower than the world average. The activity concentrations of 137Cs observed in the samples are within the range of background. concentrations. The estimated average annual effective dose from external exposure to soil and ingestion of water samples was calculated to be 0.64 mSv. The estimated outdoor external gamma dose rate measured in air ranged from 10-430 nGyh-1 with an average value of 41 nGyh-1 which is lower than the worldwide average value of 60 nGyh-1. In the case of the water samples, the average annual effective value was higher than the WHO guideline value of 0.1 mSvy-1 (author)

  2. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  3. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  4. Exergy analysis of a hydrogen fired combined cycle with natural gas reforming and membrane assisted shift reactors for CO2 capture

    International Nuclear Information System (INIS)

    Atsonios, K.; Panopoulos, K.D.; Doukelis, A.; Koumanakos, A.; Kakaras, Em.

    2012-01-01

    Highlights: ► Exergy analysis of NGCC with CCS. ► WGS-MR: exergetically efficient technology for CCS, less than 2% total exergy losses. ► 10% of total exergy dissipation in the ATR. ► Optimization of ATR operation and CO 2 stream treatment. - Abstract: Hydrogen production from fossil fuels together with carbon capture has been suggested as a means of providing a carbon free power. The paper presents a comparative exergetic analysis performed on the hydrogen production from natural gas with several combinations of reactor systems: (a) oxy or air fired autothermal reforming with subsequent water gas shift reactor and (b) membrane reactor assisted with shift catalysts. The influence of reactor temperature and pressure as well as operating parameter steam-to-carbon ratio, is also studied exergetically. The results indicate optimal power plant configurations with CO 2 capture, or hydrogen delivery for industrial applications.

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  6. Estimation of "9"9Mo production rates from natural molybdenum in research reactors

    International Nuclear Information System (INIS)

    Blaauw, M.; Chakrova, Y.; Jacimovic, R.; Kling, A.

    2017-01-01

    Molybdenum-99 is one of the most important radionuclides for medical diagnostics. In 2015, the International Atomic Energy Agency organized a round-robin exercise where the participants measured and calculated specific saturation activities achievable for the "9"8Mo(n,γ)"9"9Mo reaction. This reaction is of interest as a means to locally, and on a small scale, produce "9"9Mo from natural molybdenum. The current paper summarises a set of experimental results and reviews the methodology for calculating the corresponding saturation activities. Activation by epithermal neutrons and also epithermal neutron self-shielding are found to be of high importance in this case. (author)

  7. Predictions for heat transfer characteristics in a natural draft reactor cooling system using a second moment closure turbulence model

    International Nuclear Information System (INIS)

    Nishimura, M.; Maekawa, I.

    2004-01-01

    A numerical study is performed on the natural draft reactor cavity cooling system (RCCS). In the cooling system, buoyancy driven heated upward flow could be in the mixed convection regime that is accompanied by heat transfer impairment. Also, the heating wall condition is asymmetric with regard to the channel cross section. These flow regime and thermal boundary conditions may invalidate the use of design correlation. To precisely simulate the flow and thermal fields within the RCCS, the second moment closure turbulence model is applied. Two types of the RCCS channel geometry are selected to make a comparison: an annular duct with fins on the outer surface of the inner circular wall, and a multi-rectangular duct. The prediction shows that the local heat transfer coefficient on the RCCS with finned annular duct is less than 1/6 of that estimated with Dittus-Boelter correlation. Much portion of the natural draft airflow does not contribute cooling at all because mainstream escapes from the narrow gaps between the fins. This result and thus the finned annulus design are unacceptable from the viewpoint for structural integrity of the RCCS wall boundary. The performance of the multi-rectangular duct design is acceptable that the RCCS maximum temperature is less than 400 degree centigrade even when the flow rate is halved from the designed condition. (author)

  8. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2001-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. In the previous report, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It was found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It was also found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. However, it is very difficult to estimate the gas-liquid slip ratio theoretically, especially in the heavy liquid metal two-phase natural circulation. For example, the effects of MHD load on the two-phase flow characteristics, such as the void fraction and gas-liquid slip ratio are not known well. In the present study, therefore, as the second step of the feasibility study, a series of the experiments were performed to investigate, especially, the effect of MHD load at the single-phase shown-comer flow channel on the characteristics of the two-phase natural circulation. In the first series of the experiments, Woods-metal (Density: 9517 Kg/m 3 ) and nitrogen gas were chosen as the two-phase working fluids. The MHD pressure drop was simulated by the ball valve. The experiments with water and nitrogen gas were also performed to check the effects of the physical properties. From the present experiments, it is found that the average void fraction in the two-phase flow channel is determined by the force balance between the MHD pressure drop, frictional and pressure losses in the tube, and

  9. Natural-circulation flow pattern during the gamma-heating phase of an LBLOCA in a heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Rodriguez, S.B.; Unal, C.; Pasamehmetoglu, K.O.; Motley, F.E.

    1992-01-01

    In a postulated large-break loss-of-coolant accident (LBLOCA), the core of the reactor is uncovered quickly as the liquid that drains out of the tank is replaced by air. During the LBLOCA, the reactor is scrammed. the moderator tank is drained, and fuel and control rod tubes are cooled internally by forced convection via the emergency cooling system (ECS) water. However, the safety rods, reflector assemblies, tank wall, and instrument rods continue to heat up as a result of gamma deposition. These components are primarily cooled by natural/mixed convection and radiation heat transfer. In this paper, the thermal-hydraulic analysis of a reactor moderator tank exposed to air during an LBLOCA is discussed. The analysis was performed using a special version of the Transient Reactor Analysis Code (TRAC). TRAC input and code modifications considered the appropriate modeling of ECS cooling, thermal radiation heat transfer, and natural convection. The major objective of the model was to calculate the limiting component temperature (that establishes the maximum operating power) as a result of gamma heating. In addition, the nature of the moderator tank air-circulation pattern and its effects on the limiting temperature under various conditions were analyzed. None of the components were found to exceed their structural limits when the pre-scram power level was 50% of historical power

  10. Study on thermalhydraulics of natural circulation decay heat removal in FBR. Experiment with water of typical reactor trip in the demonstration FBR

    International Nuclear Information System (INIS)

    Koga, Tomonari; Murakami, Takahiro; Eguchi, Yuzuru

    2010-01-01

    Intending to enhance safety and to reduce costs, an FBR plant is being developed in Japan. In relies solely on natural circulation of the primary cooling loop to remove a decay heat of the core after reactor trips. A water test was carried out to advance the development. The test used a 1/10 reduced scale model simulating the core and cooling systems. The experiments simulated representative accidents from steady state to decay heat removal through reactor trip and clarified thermal-hydraulic issues on the thermal circulation performance. Some modifications of the system design were proposed for solving serious problems of natural circulation. An improved design complying with the suggestions will make it possible for natural circulation of the cooling systems to remove the decay heat of the core without causing and unstable or unpredictable change. (author)

  11. The study of capability natural uranium as fuel cycle input for long life gas cooled fast reactors with helium as coolant

    Energy Technology Data Exchange (ETDEWEB)

    Ariani, Menik, E-mail: menikariani@gmail.com; Satya, Octavianus Cakra; Monado, Fiber [Department of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University, jl Palembang-Prabumulih km 32 Indralaya OganIlir, South of Sumatera (Indonesia); Su’ud, Zaki [Nuclear and Biophysics Research Division, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, jlGanesha 10, Bandung (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, 2-12-11N1-17 Ookayama, Meguro-Ku, Tokyo (Japan)

    2016-03-11

    The objective of the present research is to assess the feasibility design of small long-life Gas Cooled Fast Reactor with helium as coolant. GCFR included in the Generation-IV reactor systems are being developed to provide sustainable energy resources that meet future energy demand in a reliable, safe, and proliferation-resistant manner. This reactor can be operated without enrichment and reprocessing forever, once it starts. To obtain the capability of consuming natural uranium as fuel cycle input modified CANDLE burn-up scheme was adopted in this system with different core design. This study has compared the core with three designs of core reactors with the same thermal power 600 MWth. The fuel composition each design was arranged by divided core into several parts of equal volume axially i.e. 6, 8 and 10 parts related to material burn-up history. The fresh natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions, i.e. shifted the core of the region (i) into region (i+1) region after the end of 10 years burn-up cycle. The calculation results shows that for the burn-up strategy on “Region-8” and “Region-10” core designs, after the reactors start-up the operation furthermore they only needs natural uranium supply to the next life operation until one period of refueling (10 years).

  12. Comparative study for axial and radial shuffling scheme effect on the performance of Pb-Bi cooled fast reactors with natural uranium as fuel cycle input

    International Nuclear Information System (INIS)

    Zaki Suud; Indah Rosidah; Maryam Afifah; Ferhat Aziz; Sekimoto, H.

    2013-01-01

    Full text:Comparative study for the Design of Pb-Bi cooled fast reactors with natural uranium as fuel cycle input using special radial shuffling strategy and axial direction modified CANDLE burn-up scheme has been performed. The reactors utilizes UN-PuN as fuel, Eutectic Pb-Bi as coolant, and can be operated without refueling for 10 years in each batch. Reactor design optimization is performed to utilize natural uranium as fuel cycle input. This reactor subdivided into 6-10 regions with equal volume in radial directions. The natural uranium is initially put in region 1, and after one cycle of 10 years of burn-up it is shifted to region 2 and the region 1 is filled by fresh natural uranium fuel. This concept is basically applied to all regions. The calculation has been done by using SRAC-Citation system code and JENDL-3.2 library. The effective multiplication factor change increases monotonously during 10 years reactor operation time. There is significant power distribution change in the central part of the core during the BOC and the EOC in the radial shuffling system. It is larger than that in the case of modified CANDLE case which use axial direction burning region move. The burn-up level of fuel is slowly grows during the first 15 years but then grow faster in the rest of burn-up history. This pattern is a little bit different from the case of modified CANDLE burn-up scheme in Axial direction in which the slow growing burn-up period is relatively longer almost half of the burn-up history. (author)

  13. Radiation exposure in the West Germany as a result of the Chernobyl reactor accident in comparison with the natural and the anthropogenic radiation exposure

    Energy Technology Data Exchange (ETDEWEB)

    Oberhausen, E

    1986-01-01

    Taking the natural radiation exposure in West Germany to be between 1 mSv (100 mrem) and 6 mSv (600 mrem), the radiation exposure due to the Chernobyl reactor accident is assessed to be in the range of 10% of natural exposure in the first year after the accident. The dose commitment assessed for the 50-year post-accident period is about 1% of natural exposure. There are no epidemiological studies available that could give information on a possible or probable increase of the individual risk to develop late damage such as cancer or genetic observations due to these very low radiation doses. (orig./HSCH).

  14. Performance evaluation of an anaerobic fluidized bed reactor with natural zeolite as support material when treating high-strength distillery wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, N. [Renewable Energy Technology Center (CETER), ' ' Jose Antonio Echeverria' ' Polytechnical University, Calle 127 s/n, CP 19390, Apdo. 6028, Habana 6 Marianao, Ciudad de La Habana (Cuba); Montalvo, S. [Department of Chemical Engineering, Santiago de Chile University, Ave. Lib. Bernardo O' Higgins 3363, Santiago de Chile (Chile); Borja, R.; Travieso, L.; Raposo, F. [Instituto de la Grasa (CSIC), Avenida Padre Garcia Tejero 4, 41012 Sevilla (Spain); Guerrero, L. [Department of Chemical, Biotechnological and Environmental Processes, Federico Santa Maria Technical University, Casilla 110-V, Valparaiso (Chile); Sanchez, E.; Colmenarejo, M.F. [Centro de Ciencias Medioambientales (CSIC), C/Serrano, 115-Duplicado, 28006 Madrid (Spain); Cortes, I. [Environment Nacional Center, Chile University, Ave. Larrain 9975, La Reina, Santiago de Chile (Chile)

    2008-11-15

    The performance of two laboratory-scale fluidized bed reactors with natural zeolite as support material when treating high-strength distillery wastewater was assessed. Two sets of experiments were carried out. In the first experimental set, the influences of the organic loading rate (OLR), the fluidization level (FL) and the particle diameter of the natural zeolite (D{sub P}) were evaluated. This experimental set was carried out at an OLR from 2 to 5 g COD (chemical oxygen demand)/l d, at FL 20% and 40% and with D{sub P} in the range of 0.2-0.5 mm (reactor 1) and of 0.5-0.8 mm (reactor 2). It was demonstrated that OLR and FL had a slight influence on COD removal, whereas they had a strong influence on the methane production rate. The COD removal was slightly higher for the highest particle diameter used. The second experimental set was carried out at an OLR from 3 to 20 g COD/l d with 25% of fluidization and D{sub P} in the above-mentioned ranges for reactors 1 and 2. The performance of the two reactors was similar; no significant differences were found. The COD removal efficiency correlated with the OLR based on a straight line. COD removal efficiencies higher than 80% were achieved in both reactors without significant differences. In addition, a straight line equation with a slope of 1.74 d{sup -1} and an intercept on the y-axis equal to zero described satisfactorily the effect of the influent COD on the COD removal rate. It was also observed that both COD removal rate and methane production (Q{sub M}) increased linearly with the OLR, independently of the D{sub P} used. (author)

  15. Natural repository analogue program. Progress report, January 1-March 30, 1982

    International Nuclear Information System (INIS)

    Curtis, D.B.

    1982-06-01

    Lead and uranium isotopic abundances in rocks from the Oklo mine show large deficiencies of radiogenic lead in the mineralized regions and enormous excesses of this element outside the uraniferous zones. A fracture lined with secondary minerals and its host rock from distances as far as approx. 13 meters away contain lead that was deposited contemporaneously. The isotopic composition of lead in these samples varies systematically as a function of distance from the fracture. This regularity may reflect the nature of the processes that transported lead from the ores and deposited it in the surrounding rocks

  16. Distributed secondary gas injection via a fractal injector : A nature-inspired approach to improving conversion in fluidized bed reactors

    NARCIS (Netherlands)

    Christensen, D.O.

    2008-01-01

    The conversion in bubbling fluidized bed reactors is suppressed because the interphase mass transfer and gas-solid contact in bubbling fluidized bed reactors are often poor. Most of the gas is present in the form of bubbles, which have low surface-to-volume ratios and are nearly devoid of catalyst

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  18. Evaluation of Radiological Impacts on the Operating Kartini Reactor and Natural Radioactivity of the Site Plan of Nuclear Power Plant Area

    International Nuclear Information System (INIS)

    Yazid, M; Sutresna, G; Sulistyono, A; Ngasifudin

    1996-01-01

    This radiological impacts evaluation covered of radioactivity in water, soil, grass, air samples and ambient gamma radiation that have been carried out in the Kartini reactor area and in the site plan of nuclear power plan are at Ujung Lemah Abang, Jepara, Central Java. The aim of this research was to determine that radiological impacts in the environment around the Kartini reactor compared to natural radioactivity for site plan of nuclear power plan area. The radioactivity in the water, soil and grass samples ware measured by low background beta counting system and were identified by low background gamma spectrometer. The radioactivity in the air samples was measured by beta portable counting system and the ambient gamma radiation was measured by portable high pressurized ionization chamber model RSS-112 Reuther-Stokes. The reactor data measurement was compared to the site plan of nuclear power plant area data for evaluation of radiological impacts on the operating reactor. From the evaluation and comparison can be concluded there are no indication of the radionuclide release from the reactor operation. The average radiactivity in the water, soil grass and air sample from the reactor area were between 0.17 - 0.61 Bq/1; 0,47 - 0,74 Bq/g; 4.43 - 4.60 Bq/g.ash and 49.53 - 70.90 x 10 Bq/cc. The average radioactivity of those sample from the nuclear power plant area were between 0.06-0.90 Bq/I; 0.02-0.86 Bq/g; 1.68-8.07 Bq/g.ash and 65.0-152.3 x 10 Bq/cc. The ambient gamma radiation were between 6.9-36.7 urad/h for the reactor area and 6.8-19.2 urad/h for the nuclear power plant area

  19. Effects of Relative SG Tube Pitches on the Performance Characteristics of a Small Modular Reactor driven by Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Youngjin; Yi, Kunwoo; Lee, Byungjin [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

    2016-10-15

    In this research, the capacity and basic dimensions for SMRs driven by a natural circulation are preliminarily assumed to determine the SMR configuration for the conceptual design, and each of the pre-set values is explained below. Firstly, the PZR configuration is not considered because it is not included to the main flow of the primary coolant. One of the SMR requirements is that SMR shall carry on the road. Hence, the vehicle geometrical limits are 15 meters for the length, and 3.5 meters for the height, approximately. With these limits for the dimensions of the SMR, RV length is assumed about 13.8 meters and RV diameter about 2.5 meters. In IAEA definition for SMRs, the capacity of electric power is no more than 300 MWe. If the efficiency of SMR power plant is assumed to 33% compared to the commercial power plant, the core power is below 1,000 MWth. In this research, the core power is assumed to 200 MWth arbitrarily during normal operation. The primary coolant passes through the outside of tubes, and the heat is transfer to the secondary feedwater. The secondary feedwater passes through the inside of tubes, and the heat from the primary coolant is received to generate the superheated steam. The present work carries out numerical simulations to get an insight for the effects of the diameters of the reactor vessel and riser using the parameters such as the steam generator tube pitches. To sum up, the calculation results show a good agreement with the theoretical equation and the uniform diameter loop has a more uniform temperature distribution and larger mass flow rate.

  20. Geochemical constraints on accumulation of actinide critical masses from stored nuclear waste in natural rock repositories. Technical report, April 1, 1978--August 31, 1978 (plus supplemental time to December 31, 1978)

    International Nuclear Information System (INIS)

    Brookins, D.G.

    1978-01-01

    Results of a literature search of abundant data on lanthanide and actinide individual and joint systematics are presented. Covered were several papers/reports about uranium solution chemistry, uranium deposits, a natural fission reactor, rare-earch deposits, manganese nodules, bedded and dome salt deposits, and miscellaneous items. This literature search is not complete but represents efforts of seven individuals attempting to gather data relevant to the objectives defined in this report. Many foreign articles, as well as many English language articles are absent. Approximately 800 articles were inspected; 69 are included in the References cited. The data search for actinides and lanthanides in natural rocks indicated that only limited segregation of the actinides U, Np, Pu, Am, and Cm from the lanthanides is possible should high-level waste be released from canisters stored in various geomedia. Supporting this were studies of Oklo and other uranium deposits, manganese nodules, monomineralic and concretion formation rates, and actinide and lathanide transport in brines. The fact that some waste canisters may, under certain conditions, contain several critical masses of one or more actinides is countered by the facts that (a) most actinides have very short half-lives and would decay before release from canisters, (b) released actinides and lanthanides, although dispersed, would be transported and deposited as a group, thus preventing point concentration of any actinides, and (c) 235 U has a much longer half-life than the other actinides, thus allowing greater time for possible reaccumulation and criticality; such a scenario would demand that 235 U be segregated effectively from other elements in the lanthanide-actinide groups.No mechanism to do this is consistent with the natural occurrences studied or the theoretical Eh-pH diagrams considered

  1. Research of natural resources saving by design studies of Pressurized Light Water Reactors and High Conversion PWR cores with mixed oxide fuels composed of thorium/uranium/plutonium

    International Nuclear Information System (INIS)

    Vallet, V.

    2012-01-01

    Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3. generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium/uranium/plutonium, and in the other hand, to elaborate multi-recycling strategies of both plutonium and 233 U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium/uranium/plutonium, and secondly the setting up of multi-recycling strategies of both plutonium and 233 U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO 2 fuel, and the second is with standard or high MR and ThUO 2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233 U production and low plutonium consumption. This second step came up with two specific concepts, from which multi-recycling strategies have been elaborated. The exclusive production and recycling of 233 U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233 U production is the limiting factor. That is why it was eventually proposed to study how the production of 233 U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233 U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%. (author) [fr

  2. Thermal-hydraulic analysis for the LBE-cooled natural circulation reactor. Development of the MSG-COPD code and application to the system analysis. Research Document

    International Nuclear Information System (INIS)

    Iwasaki, Takashi; Sakai, Takaaki; Enuma, Yasuhiro; Mizuno, Tomoyasu

    2002-03-01

    Thermal-hydraulic analysis for the Lead-Bismuth eutectic (LBE)-cooled natural circulation reactor has been conducted by using a combined plant dynamics code (MSG-COPD). MSG-COPD has been developed to consider the multi-dimensional thermal-hydraulics effect on the plant dynamics during transients. Plant dynamics analyses for the LBE-cooled STAR-LM reactor, which has been designed by Argonne National Laboratory in U.S.A., have been performed to understand the basic thermal-hydraulic characteristics of the natural circulation reactor. As a result, it has been made clear that cold coolant remains in the lower plenum by the thermal stratification in case of the ULOHS condition with a severe temperature gradient at the stratified surface in the lower plenum. In addition, the flow-redistribution effect in a core channels by the buoyancy force has been evaluated for a candidate LBE-cooled FBR plant concept (LBE-FR), which has been designed by JNC. A linear evaluation method for the flow-redistribution coefficient is proposed for the LBE-FR, and compared with the multi-dimensional results by MSG-COPD. In conclusion, the method shows sufficient performance for the prediction of the flow-redistribution coefficient for typical lateral power distributions in the core. (author)

  3. Development of core hot spot evaluation method for decay heat removal by natural circulation under transient conditions in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu; Ohkubo, Yoshiyuki

    2010-01-01

    Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot under transient conditions from forced to natural circulation operations that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal, e.g., flow redistribution in fuel assemblies caused by buoyancy force, and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional 1D and detailed 3D simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness. (author)

  4. Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

    International Nuclear Information System (INIS)

    Kamide, H.; Hayashi, K.; Momoi, K.

    1997-01-01

    The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

  5. Activación del topacio natural irradiado por neutrones en el núcleo del reactor RP-10

    OpenAIRE

    Gómez, J.; Parreño, Fernando; Lázaro, Gerardo; Vela, Mariano

    2003-01-01

    Se obtuvieron cristales de topacio activados al ser irradiados con neutrones dentro del núcleo del reactor RP-10. La activación depende del flujo de neutrones, por ello se desarrolló portamuestras (canes de irradiación) para absorber que son los causantes de la activación

  6. Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

    Energy Technology Data Exchange (ETDEWEB)

    Kruijf, W.J.M. de E-mail: kruijf@iri.tudelft.nl; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C

    2003-04-01

    Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs.

  7. Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C.

    2003-01-01

    Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs

  8. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  9. Nuclear waste criticality analysis. Quarterly progress report, 1 October--31 December 1995

    International Nuclear Information System (INIS)

    Culbreth, W.G.

    1996-01-01

    The work to date includes the preparation of a report related to criticality in spent fuel, a report on the Oklo reactors and their relevance to Yucca Mountain, and the creation of a computer program to model the Oklo reactors. The objective of the program includes a computational model of the only known natural analogue to an underground nuclear waste repository and the possible application of the model to predict the long-term behavior of Yucca Mountain. A final summary of all work completed will be presented after the end of the project on February 29, 1996

  10. Could radioactivity have played a major role in the evolution of life?

    International Nuclear Information System (INIS)

    Vendryes, G.

    2010-01-01

    A recent discovery shows the presence of the remains of multicellular organisms in rocks dating back 2.1*10 9 years which is far older than acknowledged till now. The discovery took place in Gabon near the Oklo natural reactor. The author raises the hypothesis that the passage from unicellular to multicellular organisms may have been triggered by the radiation released by uranium which was present in water at that time and began to accumulate in the sediment layers that will form the basis of the Oklo reactor millions of years later. (A.C.)

  11. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  12. natural

    Directory of Open Access Journals (Sweden)

    Elías Gómez Macías

    2006-01-01

    Full Text Available Partiendo de óxido de magnesio comercial se preparó una suspensión acuosa, la cual se secó y calcinó para conferirle estabilidad térmica. El material, tanto fresco como usado, se caracterizó mediante DRX, área superficial BET y SEM-EPMA. El catalizador mostró una matriz de MgO tipo periclasa con CaO en la superficie. Las pruebas de actividad catalítica se efectuaron en lecho fijo empacado con partículas obtenidas mediante prensado, trituración y clasificación del material. El flujo de reactivos consistió en mezclas gas natural-aire por debajo del límite inferior de inflamabilidad. Para diferentes flujos y temperaturas de entrada de la mezcla reactiva, se midieron las concentraciones de CH4, CO2 y CO en los gases de combustión con un analizador de gases tipo infrarrojo no dispersivo (NDIR. Para alcanzar conversión total de metano se requirió aumentar la temperatura de entrada al lecho a medida que se incrementó el flujo de gases reaccionantes. Los resultados obtenidos permiten desarrollar un sistema de combustión catalítica de bajo costo con un material térmicamente estable, que promueva la alta eficiencia en la combustión de gas natural y elimine los problemas de estabilidad, seguridad y de impacto ambiental negativo inherentes a los procesos de combustión térmica convencional.

  13. Okelobondo Uranium deposit: Regional context, stratigraphy, sedimentology, tectonic and mineralization

    International Nuclear Information System (INIS)

    Ango, A.M.

    1993-01-01

    This paper describes briefly the geology of Okelobondo uranium deposit (Gabon) and gives the study prospects of natural reactor phenomenon which depends of the operating progress state. Oklo phenomenon is considered as the best natural analogue for the study of radionuclide migration. 3 figs

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  15. Single and two-phase natural circulation in Westinghouse pressurized water reactor simulators: Phenomena, analysis and scaling

    International Nuclear Information System (INIS)

    Schultz, R.R.; Chapman, J.C.; Kukita, Y.; Motley, F.E.; Stumpf, H.; Chen, Y.S.; Tasaka, K.

    1987-01-01

    Natural circulation data obtained in the 1/48 scale W four loop PWR simulator - the Large Scale Test Facility (LSTF) are discussed and summarized. Core cooling modes, the primary fluid state, the primary loop mass flow and localized natural circulation phenomena occurring in the steam generator are presented. TRAC-PF1 LSTF model (using both a 1 U-tube and a 3 U-tube steam generator model) analyses of the LSTF natural circulation data including the SG recirculation patterns are presented and compared to the data. The LSTF data are then compared to similar natural circulation data obtained in the Primarkreislaufe (PKL) and the Semiscale facilities. Based on the 1/48 to 1/1705 scaling range which exists between the facilities, the implications of these data towrard natural circulation behavior in commercial plants are briefly discussed

  16. Enhancing load-following and/or spectral shift capability in single-sparger natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1992-01-01

    This patent describes a method for obtaining load-following capability in a coiling water reactor (BWR) wherein housed within a reactor pressure vessel (RPV) is a nuclear core disposed within a shroud having a shroud head and which with the RPV defines an annulus region disposed beneath the nuclear core, an upper steam dome connected to a steam outlet in the RPV, a core upper plenum formed within the shroud head and disposed atop the nuclear core, a chimney mounted atop the shroud head and in fluid communication with the core upper plenum and with a steam separator having a skirt which is in fluid communication with the steam dome, the region outside of the chimney defining a downcomer region, there being a water level established therein under normal operation of the BWR, and the RPV containing a feedwater inlet. It comprises: disposing a single sparger connected to the feedwater inlet above the steam separator skirt bottom about the interior circumference of the RPV at an elevation at approximately the water level established during normal operation of the BWR; and adjusting the feedwater flow through the inlet and into the sparger to vary the water level to be above, at or below the elevational location of the sparger in response to load-following need

  17. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  18. Experimental Study on the Natural Circulation Characteristics in the Primary Loop of the SMART Reactor by using the VISTA Facility

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Yi, Sung-Jae; Park, Choon-Kyung; Chung, Moon-Ki

    2007-01-01

    The SMART uses a two-phase natural circulation in the PRHRS loop to remove the heat from the steam generators to the PRHRS heat exchangers, while a single phase natural circulation occurs in the primary loop to transfer the decay heat from the core to the steam generator. Natural circulation operation with a power range of 20 ∼ 25% was considered for SMART and nowadays the possibility of increasing the power level during the natural circulation operation is being investigated. Previously Park et al. performed several experiments by using the VISTA facility on the thermal-hydraulic characteristics of the PRHRS for the SMART-P, which includes a single-phase natural circulation in the primary loop. From the analysis with the TASS-SMR code it was shown that the reference temperature for the primary steam generator inlet temperature should be increased in order to compensate for the decreased core flow. To investigate the possibility of an increase of the power and reference temperature, it is necessary to get experimental data to characterize the natural circulation phenomena in the primary loop of the SMART. In this paper, the characteristics of natural circulation in the primary loop are experimentally investigated during various operational conditions by using the VISTA facility

  19. Evaluation on driving force of natural circulation in downcomer for passive residual heat removal system in JAERI passive safety reactor JPSR

    International Nuclear Information System (INIS)

    Kunii, Katsuhiko; Iwamura, Takamichi; Murao, Yoshio

    1997-01-01

    The driving-force of the natural circulation in the residual heat removal (RHR) system for the JPSR (JAERI Passive Safety Reactor) is given as a gravity force of the density difference between hotter coolant in core and upper plenum and cooler coolant in downcomer. The amount of density difference and time to achieve the enough density difference for the RHR system change directly dependent on the thermal fluid flow pattern in downcomer of annulus flow pass. The purposes of the present study are to investigate the possibilities of the followings by evaluating the three-dimensional thermal fluid flow in downcomer by numerical analysis using the STREAM code; 1) promotion of making the flow pattern uniform in downcomer by installing a baffle, 2) achievement of an enough driving-force of the natural circulation, 3) validity of one-point assumption, that is, complete mixing down-flow assumption for the three-dimensional thermal fluid flow in downcomer to evaluate the function of the passive RHR system. The following conclusions were obtained: (1) The effect of baffle on the thermal fluid flow and driving-force is little, (2) The driving-force required for natural circulation cooling can be obtained in wide range of inlet velocity even if the flow is multi-dimensional, (3) Both in initial transient stage and in steady-state, the one-point assumption can be applied to evaluate the driving-force of natural circulation in the passive RHR system. (author)

  20. The ISIS operation: Robotics repair work on the CHINON A3 natural uranium, carbon dioxide cooled, graphite moderated reactor

    International Nuclear Information System (INIS)

    Hilmoine, R.M.E.

    1989-01-01

    After describing the upper internal support structures of the CHINON A3 reactor, the problems resulting from their degradation due to corrosion and to the difficulties of the ISIS operation are presented here. The repair method is as follows: all tools and repair parts reach the working area by the feeding-pipes drilled through the 7 m thick concrete vessel surrounding the reactor core; the robots handle into the reactor, the tool heads and the repair parts which are automatically positioned and welded around the corroded structure, thus restoring the support of measurement devices. The parts are either linked together or to the existing structure by means of 2 studs of 12 mm in diameter. The different phases to sort out a problem are: in-core topography, reconforming of the full-scale mock-up with the repair area, learning on this mock-up and in-core repair. The technical specificities of the robots used are the following: they have an 11 meter long, 0.22 meter across telescopic mast with jointed arms reaching a radius of 2.7 m. Then the useful load is 70 daN and the repeatability 0.1 mm. Different tool heads can be handled by the robot: telemeter and laser reconstruction: it allows to locate the in core points and to materialize them on the mock-up by a laser crossed-beams locating technique; scouring: it cleans the corroded parts of the structures before welding; welding: it allows the parts handling and the carried studs welding; screwing; tensile test: carried out when the stud welds are defective. A high level computerized control system is organized around a central unit which calculates the displacements of robots and synchronises the actions of different tools by communicating with several local units. A 100,000 hour designing, a 200,000 hour building and assembling and a 450,000 hour operating on working area were necessary to repair 15 out of the 102 corroded structures by fitting and welding 205 repair parts. 10 figs

  1. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

    The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor

  2. Performance evaluation of a natural treatment system for small communities, composed of a UASB reactor, maturation ponds (baffled and unbaffled) and a granular rock filter in series.

    Science.gov (United States)

    Dias, D F C; Passos, R G; Rodrigues, V A J; de Matos, M P; Santos, C R S; von Sperling, M

    2018-02-01

    Post-treatment of anaerobic reactor effluent with maturation ponds is a good option for small to medium-sized communities in tropical climates. The treatment line investigated, operating in Brazil, with an equivalent capacity to treat domestic sewage from 250 inhabitants, comprised a upflow anaerobic sludge blanket reactor followed by two shallow maturation ponds (unbaffled and baffled) and a granular rock filter (decreasing grain size) in series, requiring an area of only 1.5 m 2  inhabitant -1 . With an overall hydraulic retention time of only 6.7 days, the performance was excellent for a natural treatment system. Based on over two years of continuous monitoring, median removal efficiencies were: biochemical oxygen demand = 93%, chemical oxygen demand = 79%, total suspended solids = 87%, ammonia = 43% and Escherichia coli = 6.1 log units. The final effluent complied with European discharge standards and WHO guidelines for some forms of irrigation, and appeared to be a suitable alternative for treating domestic sewage for small communities in warm areas, especially in developing countries.

  3. Analytical evaluation on dynamical response characteristics of reduced-moderation water reactor with tight-lattice core under natural circulation core cooling

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

    The time-domain analyses with TRAC-BF1 code were performed for clarifying the dynamical response characteristics of the reduced-moderation water reactor (RMWR) with tight-lattice core configuration. The response characteristics were evaluated based on the step response basically utilized for dynamical system evaluation. As for the most fundamental dynamical characteristics, the channel flow response characteristics of single fuel assembly were evaluated. In the evaluation, the appropriate single-phase pressure drop setting at the inlet orifice was determined in terms of response stability from the design viewpoint. In addition, from the investigation on the relation of the response and transit time of coolant, it is confirmed that the channel flow response of RMWR is dominated by the transit time of vapor phase resulting from a high void fraction operation condition. As for a natural circulation flow response, it is clarified that the response is strongly influenced by the effect of two-phase pressure loss owing to a high void fraction condition. The reactor power response with reactivity feedback shows quite stable response characteristics on account of the small absolute value of void reactivity coefficient.

  4. Operation of the NETL Chemical Looping Reactor with Natural Gas and a Novel Copper-Iron Material

    Energy Technology Data Exchange (ETDEWEB)

    Straub, Douglas [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Bayham, Samuel [National Energy Technology Lab. (NETL), Morgantown, WV (United States); Weber, Justin [National Energy Technology Lab. (NETL), Morgantown, WV (United States)

    2017-02-21

    The proposed Clean Power Plan requires CO2 emission reductions of 30% by 2030 and further reductions are targeted by 2050. The current strategies to achieve the 30% reduction targets do not include options for coal. However, the 2016 Annual Energy Outlook suggests that coal will continue to provide more electricity than renewable sources for many regions of the country in 2035. Therefore, cost effective options to reduce greenhouse gas emissions from fossil fuel power plants are vital in order to achieve greenhouse gas reduction targets beyond 2030. As part of the U.S. Department of Energy’s Advanced Combustion Program, the National Energy Technology Laboratory’s Research and Innovation Center (NETL R&IC) is investigating the feasibility of a novel combustion concept in which the GHG emissions can be significantly reduced. This concept involves burning fuel and air without mixing these two reactants. If this concept is technically feasible, then CO2 emissions can be significantly reduced at a much lower cost than more conventional approaches. This indirect combustion concept has been called Chemical Looping Combustion (CLC) because an intermediate material (i.e., a metal-oxide) is continuously cycled to oxidize the fuel. This CLC concept is the focus of this research and will be described in more detail in the following sections. The solid material that is used to transport oxygen is called an oxygen carrier material. The cost, durability, and performance of this material is a key issue for the CLC technology. Researchers at the NETL R&IC have developed an oxygen carrier material that consists of copper, iron, and alumina. This material has been tested extensively using lab scale instruments such as thermogravimetric analysis (TGA), scanning electron microscopy (SEM), mechanical attrition (ASTM D5757), and small fluidized bed reactor tests. This report will describe the results from a realistic, circulating, proof-of-concept test that was

  5. CFD investigations of natural circulation between the RPV and the cooling pond of VVER-440 type reactors in incidental conditions during maintenance performed with the code CFX-4.3

    International Nuclear Information System (INIS)

    Legradi, G.; Aszodi, A.

    2002-01-01

    During the annual maintenance of the VVER-440 type reactors, the RPV, the cooling pond and the transfer pond form a connected flow domain. The reactor is cooled by the natural circulation, which develops in one or two main loops. The cooling pond has its own cooling loops. CFD calculations have been performed with the CFX-4.3 code to investigate whether it is possible to cool the reactor core in case the main loops are lost and other emergency systems are not available. The results point out that the cooling system of the cooling pond is not capable of cooling the reactor core with the present connection. Therefore, modifications of the cooling system are investigated which would make it suitable for removing the remanent heat from the core.(author)

  6. Effects of post-burial siliceous diagenesis deformations on the microthermometric behaviour of fluid inclusions: an example in the Francevillian uraniferous sandstone reservoir (Gabon)

    International Nuclear Information System (INIS)

    Gauthier-Lafaye, F.; Weber, F.

    1984-01-01

    New data about fluid inclusions associated to a siliceous diagenesis show that a deformation phase in the first stage of catagenesis disturbed their microthermometric behaviour. Nevertheless, temperature and pressure of fluids associated to the uraniferous paragenesis and contemporary with the Oklo natural reactors are estimated at 140-160 0 C and 250-500 bar [fr

  7. Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.1

    International Nuclear Information System (INIS)

    2014-01-01

    The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

  8. Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.2

    International Nuclear Information System (INIS)

    2014-01-01

    The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

  9. Numerical calculation and analysis of natural convection removal of the spent fuel residual heat of 10 MW high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Wang Jinhua; Huang Yifan; Wu Bin

    2013-01-01

    The spent fuel of 10 MW High Temperature Gas Cooled Reactor (HTR-10) could be stored in the shielded tank, and the tank is stored in the concrete shielded canister in spent fuel storage room, the residual heat of the spent fuel could be removed by the air. The ability of residual heat removal is analyzed in the paper, and the temperature field is numerically calculated through FEA program ANSYS, the analysis and the calculation are used to validate the safety of the spent fuel and the tank, the ultimate temperature of the spent fuel and the tank should below the safety limit. The calculation shows that the maximum temperature locates in the middle of the fuel pebble bed in the spent fuel tank, and the temperature decreases gradually with radial distance, the temperature in the tank body is evenly distributed, and the temperature in the concrete shielded canister decreases gradually with radial distance. It is feasible to remove the residual heat of the spent fuel storage tank by natural ventilation, in natural ventilation condition, the temperature of the spent fuel and the tank is lower than the temperature limit, which provides theoretical evidence for the choice of the residual heat removal method. (authors)

  10. Household energy consumption: the future is in our hands. ITER, an experimental fusion reactor. Do CO2-free energies exist? Liquefied natural gas, king of the gas market

    International Nuclear Information System (INIS)

    Anon.

    2008-01-01

    This issue of Alternatives newsletter features 4 main articles dealing with: 1 - Household energy consumption - the future is in our hands: With energy resources growing scarcer and more expensive, everyone has a duty to conserve energy. Because combating global warming also means adopting simple habits and using the right equipment - with help from our governments to lead us to change. A practical look at what we can do. 2 - ITER, an experimental fusion reactor: The entire international community is trying to reproduce here on Earth the fusion of hydrogen atoms occurring naturally in the Sun, lured by the promise of a virtually inexhaustible source of energy. More on ITER from the project's Director General. 3 - Do CO 2 -free energies exist?: As nations struggle to reduce greenhouse gas emissions, the question is moot. Environmental engineer Jean-Marc Jancovici gives us his point of view. 4 - Liquefied natural gas, king of the gas market: LNG's many advantages are enticing industry to develop supply routes and infrastructure to meet strong demand. But the race for LNG is not without its limits

  11. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  12. Natural and artificial radionuclides in selected Styrian soils and plants before and after the reactor accident in Chernobyl

    Energy Technology Data Exchange (ETDEWEB)

    Heinrich, G; Gries, A [Graz Univ. (Austria); Mueller, H J; Oswald, K [Technische Univ., Graz (Austria). Inst. fuer Theoretische Physik und Reaktorphysik

    1989-01-01

    This paper reports on natural radioactivity due to the uptake of {sup 40}K and the radionuclides of the {sup 238}U and {sup 232}Th series. Selected examples show the concentrations of radionuclides in soils, lower and higher plants before and after the Chernobyl accident. The changes in the amount of radioactivity which have occurred during the vegetation periods of 1987 and 1988 have been investigated. It is also demonstrated how the conditions have changed in the following two periods of growth. In leaves of deciduous trees which were directly contaminated in 1986 natural radioactivity is sometimes higher than artificial one today. This has not been observed in conifers, because the needles contaminated in 1986 have not yet been shed. The {sup 137}Cs activity in mosses and lichens has hardly decreased. It is therefore possible to produce autoradiographs at the Styrian test site locations. Within the same genus of fungi, Cs-discriminating and Cs-accumulating species have been noted. In the latter, radioactivity has probably increased since 1986. (author).

  13. Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Jaehyun [Korea Atomic Energy Research Institute, 111 Daedeok-daero, 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jueun; Ju, Heejae; Sohn, Sungjune; Kim, Yeji; Noh, Hyunyub; Hwang, Il Soon [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of)

    2017-05-15

    Highlights: • Experimental studies on natural circulation for lead-bismuth eutectic were conducted. • Adiabatic wall boundaries conditions were established by compensating heat loss. • Computational benchmark with a system thermal-hydraulics code was performed. • Numerical simulation and experiment showed good agreement in mass flow rate. • An empirical relation was formulated for mass flow rate with experimental data. - Abstract: In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8 kW to 33.6 kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

  14. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  15. Concomitant formation of different nature clusters and hardening in reactor pressure vessel steels irradiated by heavy ions

    International Nuclear Information System (INIS)

    Fujii, K.; Fukuya, K.; Hojo, T.

    2013-01-01

    Specimens of A533B steels containing 0.04, 0.09 and 0.21 wt%Cu were irradiated at 290 °C to 3 dpa with 3 MeV Fe ions and subjected to atom probe analyses, transmission electron microscopy observations and hardness measurements. The atom probe analysis results showed that two types of solute clusters were formed: Cu-enriched clusters containing Mn, Ni and Si atoms as irradiation-enhanced solute atom clusters and Mn/Ni/Si-enriched clusters as irradiation-induced solute atom clusters. Both cluster types occurred in the highest Cu-content steel and the ratio of Mn/Ni/Si-enriched clusters to Cu-enriched clusters increased with irradiation doses. It was confirmed that the cluster formation was a key factor in the microstructure evolution until the high dose irradiation was reached even in the low Cu content steels though the dislocation loops with much lower density than that of the clusters were observed as matrix damage. The difference in the hardening efficiency due to the difference in the nature of the clusters was small. The irradiation-induced clustering of undersized Si atoms suggested that a clustering driving force other than vacancy-driven diffusion, probably an interstitial mechanism, may become important at higher dose rates

  16. Concomitant formation of different nature clusters and hardening in reactor pressure vessel steels irradiated by heavy ions

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, K., E-mail: fujiik@inss.co.jp [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Fukuya, K. [Institute of Nuclear Safety System, Inc., Mihama 919-1205 (Japan); Hojo, T. [Japan Nuclear Energy Safety Organization, Toranomon, Minato-ku, Tokyo 105-0001 (Japan)

    2013-11-15

    Specimens of A533B steels containing 0.04, 0.09 and 0.21 wt%Cu were irradiated at 290 °C to 3 dpa with 3 MeV Fe ions and subjected to atom probe analyses, transmission electron microscopy observations and hardness measurements. The atom probe analysis results showed that two types of solute clusters were formed: Cu-enriched clusters containing Mn, Ni and Si atoms as irradiation-enhanced solute atom clusters and Mn/Ni/Si-enriched clusters as irradiation-induced solute atom clusters. Both cluster types occurred in the highest Cu-content steel and the ratio of Mn/Ni/Si-enriched clusters to Cu-enriched clusters increased with irradiation doses. It was confirmed that the cluster formation was a key factor in the microstructure evolution until the high dose irradiation was reached even in the low Cu content steels though the dislocation loops with much lower density than that of the clusters were observed as matrix damage. The difference in the hardening efficiency due to the difference in the nature of the clusters was small. The irradiation-induced clustering of undersized Si atoms suggested that a clustering driving force other than vacancy-driven diffusion, probably an interstitial mechanism, may become important at higher dose rates.

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  18. Uncertainties and credibility building of safety analyses. Natural analogues

    International Nuclear Information System (INIS)

    Laciok, A.

    2001-07-01

    The substance of natural analogues and their studies is defined as a complementary method to laboratory and in-situ experiments and modelling. The role of natural analogues in the processes of development of repositories is defined, mainly in performance assessment of repository system and communication with public. The criteria for identification of natural analogues which should be evaluated in the phase of initiation of new studies are specified. Review part of this report is divided to study of natural analogues and study of anthropogenic and industrial analogues. The main natural analogue studies performed in various countries, in different geological setting, with various aims are characterized. New results acquired in recently finished studies are included: Palmottu (2nd phase of project financed by European Commission), Oklo (results of research financed also by European Commission), Maqarin (3rd phase) and other information obtained from last meetings and workshops of NAWG. In view of the fact that programmes of development of deep repositories in Czech and Slovak Republics are interconnected, the natural analogues studies carried out in the Czech republic are incorporated in separate chapter - study of uranium accumulation in Tertiary clays at Ruprechtov site and study of degradation of natural glasses. In final part the areas of natural analogue studies as an integral part of development of deep geological repository are proposed along with characterization of broader context and aspects of realization of these studies (international cooperation, preparation and evaluation of procedures, communication with public). (author)

  19. A multi-purpose reactor

    International Nuclear Information System (INIS)

    Changwen Ma

    2000-01-01

    An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

  20. The variation of the reactivity with the number, diameter and length of the control rods in a heavy water natural uranium reactor

    Energy Technology Data Exchange (ETDEWEB)

    McCriric, H

    1958-05-15

    Starting with the known reactor constants for a heavy water moderated reactor with reflector and a given number of control rods of a certain size, the reactivity equivalence of the control rods is calculated. The calculation is given in detail. The number, length and diameter of the control rods is then varied and the effect of these parameters on the reactivity is shown graphically. Flux plots are also given for the reactor with and without control rods.

  1. Physical and numerical modelling of natural convection in a fluid layer of small aspect ratio, in the frame of severe accidents of nuclear pressurized water reactors

    International Nuclear Information System (INIS)

    Villermaux, Clotilde

    1999-01-01

    In the framework of PWR reactor accidents studies, the possibility of cooling the corium by the vessel flooding, is analysed. A particular attention is given to the liquid materials of the upper part of this pool. The confinement and the physical properties of this melt pool, may threat the vessel integrity by a heat flux concentration on the vessel lateral wall. A bibliographic study on the thermal transfers in natural convection, enhances the influence of the thermal extreme conditions and the layer geometry on the flow structure and the heat distribution. The lower part of the corium is constituted of an oxides layer. A stability study shows its perenniality: the metallic layer can be slipped of the oxides pool. The results analysis of the experimental program, BALI-metal, is completed by a direct numerical simulation with the TRIOU code. A model of the flow structure allows the find in bulk the experimental results. Finally a numerical simulation of the experimental tests is realized with the thermo-hydraulic code TOLBIAC. (A.L.B.)

  2. Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

    International Nuclear Information System (INIS)

    Loktionov, V.D.; Mukhtarov, E.S.; Yaroshenko, N.I.; Orlov, V.E.

    1999-01-01

    Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of 'overpressure-molten pool structure' when the vessel failure started at the 'hot' layers of the vessel. (orig.)

  3. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  4. Definition of validated membrane reactor model for 5 kW power output CHP system for different natural gas compositions

    NARCIS (Netherlands)

    Di Marcoberardino, Gioele; Gallucci, Fausto; Manzolini, Giampaolo; van Sint Annaland, Martin

    2016-01-01

    Over the last years, many studies focused on the development of membrane reactors for micro-cogeneration systems based on PEM fuel cells, thanks to its unique feature of separating pure hydrogen. This work deals with (i) the design of a fluidized bed membrane reactor flexible towards different

  5. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  6. The dissolution kinetics and apparent solubility of natural apatite in closed reactors at temperatures from 5 to 50 degrees C and pH from 1 to 6

    Energy Technology Data Exchange (ETDEWEB)

    Harouiya, N.; Chairat, C.; Kohler, S.J.; Gout, R.; Oelkers, E.H. [Univ Toulouse 3, CNRS, UMR 5563, F-31400 Toulouse (France); Chairat, C. [CEA, LCLT SECM DTCD, Lab Etud Comportement Long Terme, F-30207 Bagnols Sur Ceze, (France)

    2007-07-01

    The apparent solubility and dissolution rates of natural apatite were measured in closed-system reactors as a function of temperature from 5 to 50 degrees C and pH from 1 to 6. The temporal release rates of Ca, P, and F during the experiments are approximately consistent with stoichiometric dissolution in all experiments. One advantage of closed-system experiments is that they allow determination of reactive fluid evolution and dissolution rates at far-from to near-to equilibrium conditions. Surface area normalized apatite dissolution rates, r, obtained in all experiments are consistent with r = A{sub A}a{sub H{sup +}}{sup n}exp(E{sub A}/RT)(1 -exp(-A/{sigma} RT)) where A{sub A} stands for a rate constant equal to 4 * 10{sup -3} mol/cm{sup 2}/s, a{sub H{sup +}}) denotes the activity of the aqueous H{sup +}, n designates a reaction order equal to 0.6, E{sub A} symbolizes an activation energy equal to 11.0 kcal/mol, A refers to the chemical affinity of the dissolving apatite, {sigma} stands for Temkin's average stoichiometric number equal to 5; R designates the gas constant, and T represents absolute temperature. Logarithms of apparent equilibrium constants obtained from experiments performed at 3 {<=} pH {<=} 5.6 for the apatite dissolution reaction: Ca{sub 5}(PO{sub 4}){sub 3}F + 3H{sup +} = 5Ca{sup 2+} + 3HPO{sub 4}{sup 2-} + F{sup -} are found to be - 29.5 {+-} 0.6, - 29.4 {+-} 0.9 and - 29.9 {+-} 1.3 at 5, 25, and 50 degrees C, respectively. (authors)

  7. From a critical assembly heavy water - natural uranium to the fast - thermal research reactor in the Institute Vinca; Od kriticnog sistema teska voda - prirodni uranium do brzo - termickog istrazivackog reaktora u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovic, D; Pesic, M [Vinca Institute of Nuclear Sciences, Beograd (Yugoslavia)

    1995-07-01

    A part of the Institute in Vinca this monograph refers to is the thermal nuclear zero power reactor RB, with a heavy water moderator and variously enriched uranium fuel, that is, its present day version, the coupled fast-thermal system HERBE. A group of research workers, technicians, operators and skilled workmen in the workshop have worked continuously on it. Some of them have spent their whole working age at the reactor, and some a part of it. There is about a hundred and fifty internationally published papers, twenty master's and fourteen doctor's theses left behind them for the past thirty five years. This book is devoted to them. The first part of the text refers to the pioneering efforts on the reactor and fundamental research in reactor physics. The experimental reactor RB was designed and constructed at the time to operate with natural uranium and heavy water. Measurements are presented and the first results of reaching critical state, measurements of migration length of thermal neutrons and neutron multiplication factor in an infinite medium; also measurements of neutron flux density distribution and reactor parameter, and in the domain of safety, measurement of safety rods reactivity. Those were also the times when the known serious accident occurred with the uncontrolled rise of reactivity, which was especially minutely described in a publication of the International Atomic Energy Agency from Vienna. Later on, new fuel was acquired with 2 % enriched uranium. A series of experiments in reactor and neutron physics followed, with just the most interesting results of them presented here. In the period which followed, another type of fuel was available, with 80 % enriched uranium. New possibilities for work opened. Measurements with mixed lattices were performed, and the RA reactor lattices were simulated. After measurements mainly in the sphere of reactor and neutron physics, a need for investigations in the field of gamma and neutron radiation protection

  8. Anomalous deuteron to hydrogen ratio in naturally occuring fission reactions and the possibility of deuteron disintegration

    International Nuclear Information System (INIS)

    Shaheen, M.; Ragheb, M.

    1992-01-01

    A hypothesis is presented for explaining the experimentally determined anomalous D/H ratio observed in the samples from the naturally occuring fission reaction in the Oklo phenomenon. No other explanation has been given, to the best knowledge, for the large difference between the measured D/H ratio in the Oklo samples and the expected values in a fission neutron spectrum. A multicomponent system consisting of hydrogen, deuterium, tritium and helium nuclei is considered. An analytical solution is derived and solved using as boundary conditions the experimentally determined value of the D/H ratio. The solution of the rate equations for hydrogen and deuteron concentrations, assuming a pure fission process without a deuteron sink term, yields a D/H ratio of 445 ppm for a reaction in which the fluence of neutrons is 10 21 n/cm 2 . This exceeds the experimentally observed value of 127 ppm, and the naturally occuring value of 150 ppm. Solving the same rate equations accounting for a deuterium sink term using a hypothesis of deuteron disintegration, and the experimentally observed value of 127 ppm yields a deuteron disintegration constant of 7.47*10 -14 s -1 . Deuteron disintegration would provide a neutron source, in addition to the fission neutrons, driving a subcritical chain reaction over an extended period of time. Relationship of the presented hypothesis to the Vlasov theory of an annihilation meteorite impact explosion explaining the experimentally observed anomalous 235 U/ 238 U ratio, and to the suggestion of deuteron disintegration as a possible explanation of some observations of deuterium dissociation in palladium and titanium electrodes is discussed. The tritium andhelium-3 rate equations are further solved under the deuteron disintegration hypothesis and the relationship of the present work to the work by JONES et al. is discussed. (author) 16 refs.; 7 figs.; 2 tabs

  9. NEUTRONIC REACTOR STRUCTURE

    Science.gov (United States)

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  10. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

    DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

  11. Definition of parameters for a test section for the analysis of natural convection and coolant loss in the AP1000 nuclear reactor by similarity laws and fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Cadiz, Luis Felipe S.; Bezerra, Mario Augusto [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear; Lima, Fernando Roberto A., E-mail: falima@cnen.gov.br [Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE/CNEN-PB), Recife, PB (Brazil)

    2017-07-01

    The present work develops and analyzes the main parameters of a test section for natural convection in case of a failure of the pumping system as much as the loss of coolant in refrigeration accidents. For this realization, a combination of laws of basic similarity and an innovative scale methodology, known as Fractional Scaling Analysis (FSA), was developed. The depressurizing is analyzed when a rupture occurs in one of the primary system piping of the AP1000 nuclear reactor. This reactor is developed by Westinghouse Electric Co., which is a PWR (Pressurized Water Reactor) with an electric power equal to 1000MW. Such a reactor is provided with a passive safety system that promotes considerable improvements in the safety, reliability, protection and reduction of costs of a nuclear power plant. The FSA is based on two concepts: fractional scale and hierarchy. It is used to provide experimental data that generate quantitative evaluation criteria as well as operational parameters in thermal and hydraulic processes of nuclear power plants. The results were analyzed with the use of computational codes. (author)

  12. Inherently safe reactors

    International Nuclear Information System (INIS)

    Maartensson, Anders

    1992-01-01

    A rethinking of nuclear reactor safety has created proposals for new designs based on inherent and passive safety principles. Diverging interpretations of these concepts can be found. This article reviews the key features of proposed advanced power reactors. An evaluation is made of the degree of inherent safety for four different designs: the AP-600, the PIUS, the MHTGR and the PRISM. The inherent hazards of today's most common reactor principles are used as reference for the evaluation. It is concluded that claims for the new designs being inherently, naturally or passively safe are not substantiated by experience. (author)

  13. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  14. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  15. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  16. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  17. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  18. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  20. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  1. Multi-purpose reactor

    International Nuclear Information System (INIS)

    1991-05-01

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

  2. Mirror hybrid reactor optimization studies

    International Nuclear Information System (INIS)

    Bender, D.J.

    1976-01-01

    A system model of the mirror hybrid reactor has been developed. The major components of the model include (1) the reactor description, (2) a capital cost analysis, (3) various fuel management schemes, and (4) an economic analysis that includes the hybrid plus its associated fission burner reactors. The results presented describe the optimization of the mirror hybrid reactor, the objective being to minimize the cost of electricity from the hybrid fission-burner reactor complex. We have examined hybrid reactors with two types of blankets, one containing natural uranium, the other thorium. The major difference between the two optimized reactors is that the uranium hybrid is a significant net electrical power producer, whereas the thorium hybrid just about breaks even on electrical power. Our projected costs for fissile fuel production are approximately 50 $/g for 239 Pu and approximately 125 $/g for 233 U

  3. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  4. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  6. Report of the Federal Ministry for the Environment, Protection of Nature and Reactor Safety, on the reactor accident of Chernobyl, its repercussions, and precautions taken or to be taken - including addenda

    International Nuclear Information System (INIS)

    Petroll, M.

    1986-01-01

    Apart from the report of the Federal Ministry for the Environment, the publication contains the following chapters: 1) Monitoring of environmental radioactivity; 2) analysis of propagation processes; 3) control and measuring points of the Federal Laender to monitor environmental radioactivity; 4) determination of the local dose rate; 5) concentration of radioactivity in air and soil (graphs); 6) up-to-date knowledge of events, measures; 7) nuclear power plants in the Federal Republic of Germany (review of technical safety); 8) interim report of the Committee on Reactor Safety - Reaktorsicherheitskommission - for preliminary evaluation; 9) interim report of the Committee on Radiation Protection - Strahlenschutzkommission - for assessment and evaluation of the effects. (HP) [de

  7. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  8. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  9. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  10. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  11. Theoretical Calculations of the Effect on Lattice Parameters of Emptying the Coolant Channels in a D{sub 2}O- Moderated and Cooled Natural Uranium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

    1960-11-15

    The purpose of the present study was to evaluate theoretically the effect of coolant boiling and subsequent void formation in a pressurized D{sub 2}O moderated and cooled reactor. The fuel rods were arranged in a cluster geometry and clad in Zr-2. The coolant was separated from the moderator by a Zr-2 shroud. In this geometry the following problems have been given special attention: l) calculation of the effective resonance integral, 2) thermal disadvantage factors, 3) fast fission effects, 4) leakage effects, 5) changes in epithermal absorption. No account has up to now been taken of the variation of these effects with position in the reactor and burnup. Some comparisons of the theoretical methods and measurements have been attempted. It is concluded that at the present time it is not possible to calculate the void coefficient with any accuracy but it may be possible to give an upper limit from theoretical consideration.

  12. 2nd RCM of the CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The overall objective of the CRP is to improve the Member States’ analytical capabilities in the field of fast reactor in-vessel sodium thermal hydraulics. A necessary condition towards achieving this objective is a wide international validation effort of the data and codes currently employed for the simulation of the various physical effects involved in this field. Therefore, in providing the required wide international basis of interested Member States, each applying different methodologies, the CRP will contribute towards achieving the stated objective with the help of benchmark exercises focusing, in a first stage, on the numerical simulation of temperature stratification of sodium observed in the Monju reactor vessel at a turbine trip test conducted in December 1995 during the original start-up experiments, and with the help of a thorough assessment of the calculation versus measured data comparisons

  13. Device for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yanagisawa, Yutaro; Kawarazaki, Yuki; Sugiyama, Yu.

    1996-01-01

    A member comprising hydrogen occluding materials is introduced to a reactor incorporated with U-235 as fuels in order to moderate and breed fast neutrons and to control the reactor. Since the amount of light hydrogen or heavy hydrogen is substantially the same as that of metal, etc. of hydrogen occluding material, a moderating efficiency substantially equal with that of a moderator comprising H 2 O can be obtained. In addition, since the member acting as a moderator has an effect of multiplying neutrons, use of only natural uranium 0.72% as nuclear fuels causes chain reaction to provide a function as a nuclear reactor. Further, the hydrogen occluding material can be used also as a control rod for controlling the reactor. The hydrogen occluding material may be Ti, Zr, Pd, proton conductor, Ag, Pt, Rh or oxides thereof or alloys thereof. The member comprising hydrogen occluding materials is preferably coated with a material not permeating hydrogen. (N.H.)

  14. Surveys of research projects concerning nuclear facility safety, financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1988

    International Nuclear Information System (INIS)

    1989-11-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  15. Surveys of research projects concerning nuclear facility safety, financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1987

    International Nuclear Information System (INIS)

    1988-06-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  16. Surveys of research projects concerning nuclear facility safety financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1991

    International Nuclear Information System (INIS)

    1992-09-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  17. Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

    International Nuclear Information System (INIS)

    Zhao, Hangbin; Yan, Changqi; Sun, Licheng; Zhao, Kaibin; Fa, Dan

    2015-01-01

    As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

  18. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  19. Determination of the isotopic abundance of 235U in rocks in search for an Oklo phenomenon in Brazil by activation analysis

    International Nuclear Information System (INIS)

    Vasconcellos, M.B.A.; Armelin, M.J.A.; Lima, F.W. de; Fulfaro, R.

    1981-09-01

    Isotopic analyses of uranium are generally carried out by mass spectrometry, with a precision better than 1%. In nuclear laboratories it is often necessary to perform rapid determinations of 235 U isotopic abundances. Thermal neutron activation analysis by delayed neutron counting or by high resolution gamma-ray spectrometry can be applied for this purpose, although with less precision than by mass spectrometry. In this work, delayed neutron counting and gamma-ray spectrometry are used for the determination of the isotopic abundance of 235 U in rocks from the Northeastern region of Brazil. In the case of the application of delayed neutron counting, the rocks are analyzed non-destructively. When high resolution gamma-ray spectrometry is applied, a pre-irradiation chemical separation had to be performed, by extraction of uranium with tributylphosphate. By both methods employed the results for the isotopic abundance of 235 U can be considered as equal to the natural value of 0.702%, for the rocks under study. The precision attained by gamma-ray spectrometry is better than that by delayed neutron couting. (Author) [pt

  20. Dating of the Francevillian sedimentary series and mineralogic and isotopic (Sm, Nd, Rb, Sr, K, Ar, U, O and C) characterization of the gangue of the reactors 10 and 13. Preliminary report

    International Nuclear Information System (INIS)

    Gautier-Lafaye, F.; Stille, P.; Bros, R.; Taieb, R.

    1993-01-01

    This paper summarizes the various ages reported for the diagenetic events in the Francevillian sedimentary series (Precambrian era) and the fission reactors of Oklo. Obviously, differences exist between the ages obtained on the silicate minerals and the ages obtained on the Uranium ores and on the reactors. Clay minerals which crystallized during the fission reactions yield younger ages than the reactors themselves. Similarly, the diagenetic clays (1870 Ma) show younger ages than the Uranium ores (2000 Ma). This is in contrast to mineralogical and field evidence indicating that Uranium mineralization occurred during diagenesis of the Francevillian sediments. These antithetical results give rise to several questions. Does the age obtained on the diagenetic clays date a late thermal event or does the age of the Uranium mineralization reflect a multistage U-Pb history. This work tries to bring answers with the help of new isotopic analysis and studies mineralogy of the gangue of reactors and isotopic compositions in Uranium ores. 8 refs., 4 figs

  1. Advanced gas-cooled reactors (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

    1981-01-15

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

  2. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  3. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  5. Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France; Quelques aspects economiques de la filiere uranium naturel - Graphite - gaz. Etat actuel et tendance des couts en France

    Energy Technology Data Exchange (ETDEWEB)

    Gaussens, J; Tanguy, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leo, B [Electricite de France (EDF), 75 - Paris (France)

    1964-07-01

    The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable

  6. Radio-active pollution near natural uranium-graphite-gas reactors; La pollution radioactive aupres des piles uranium naturel - graphite - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J.; Pouthier, J.; Delmar, J. [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1967-07-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [French] Les resultats de nombreuses evaluations de la contamination sont donnes: - Piles en marche pendant les operations d'entretien - Piles a l'arret au cours des chantiers caracteristiques (refrigerants, echangeurs, interieur du caisson, etc.) - A la suite d'incidents sur le circuit de refroidissement et de rupture de gaine. Ils montrent que, sauf cas particulier, ce sont essentiellement les produits d'activation qui dominent. Par ailleurs apres 10 ans de fonctionnement, les points d'accumulation de la contamination susceptibles de delivrer des debits de dose importants restent tres localises et les moyens de protection individuels utilises n'ont pas du etre renforces. (auteurs)

  7. Radio-active pollution near natural uranium-graphite-gas reactors; La pollution radioactive aupres des piles uranium naturel - graphite - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Chassany, J; Pouthier, J; Delmar, J [Commissariat a l' Energie Atomique, Chusclan (France). Centre de Production de Plutonium de Marcoule

    1967-07-01

    The results of numerous evaluations of the contamination are given: - Reactors in operation during maintenance operations. - Reactors shut-down during typical repair operations (coolants, exchangers, interior of the vessel, etc. ) - Following incidents on the cooling circuit and can-rupture. They show that, except in particular cases, it is the activation products which dominate. Furthermore, after ten years operation, the points at which contamination liable to emit strong doses accumulates are very localized and the individual protective equipment has not had to be reinforced. (authors) [French] Les resultats de nombreuses evaluations de la contamination sont donnes: - Piles en marche pendant les operations d'entretien - Piles a l'arret au cours des chantiers caracteristiques (refrigerants, echangeurs, interieur du caisson, etc.) - A la suite d'incidents sur le circuit de refroidissement et de rupture de gaine. Ils montrent que, sauf cas particulier, ce sont essentiellement les produits d'activation qui dominent. Par ailleurs apres 10 ans de fonctionnement, les points d'accumulation de la contamination susceptibles de delivrer des debits de dose importants restent tres localises et les moyens de protection individuels utilises n'ont pas du etre renforces. (auteurs)

  8. EXTRAPOLATING THE SUITABILITY OF SOILS AS NATURAL REACTORS USING AN EXISTING SOIL MAP: APPLICATION OF PEDOTRANSFER FUNCTIONS, SPATIAL INTEGRATION AND VALIDATION PROCEDURES

    Directory of Open Access Journals (Sweden)

    Yameli Guadalupe Aguilar Duarte

    2011-04-01

    Full Text Available The aim of this study was the spatial identification of the suitability of soils as reactors in the treatment of swine wastewater in the Mexican state of Yucatan, as well as the development of a map with validation procedures. Pedotransfer functions were applied to the existing soils database. A methodological approach was adopted that allowed the spatialization of pedotransfer function data points. A map of the suitability of soil associations as reactors was produced, as well as a map of the level of accuracy of the associations using numerical classification technique, such as discriminant analysis. Soils with the highest suitability indices were found to be Vertisols, Stagnosols, Nitisols and Luvisols. Some 83.9% of the area of Yucatan is marginally suitable for the reception of swine wastewater, 6.5% is moderately suitable, while 6% is suitable. The percentages of the spatial accuracy of the pedotransfer functions range from 62% to 95% with an overall value of 71.5%. The methodological approach proved to be practical, accurate and inexpensive.

  9. Natural uranium-graphite system. Critial experiments on the G1 reactor; Systeme uranium naturel-graphite. Experiences critiques sur le reacteur G1

    Energy Technology Data Exchange (ETDEWEB)

    Schmitt, A P; Tanguy, P; Teste du Bailler, A; Zaleski, C P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    A number of experiments have been performed during the start up period of the G1 (1956) and G2 (1958) reactors in Marcoule, both on their lattices and on different lattices (hollow rods, clusters, under moderated lattices). The first chapter gives a thorough description of the two reactors. The second chapter deals with buckling measurements, both absolute (flux plots) and relative by the method of progressive substitution. The experimental results are summarised in Table VI. The third chapter contains a number of other measurements performed on G1. (author)Fren. [French] Le demarrage des reacteurs G1 (1956) et G2 (1958) de Marcoule nous a permis d'effectuer une serie d'experiences tant sur les reseaux de ces piles que sur des reseaux differents (elements tubulaires ou divises, reseaux sous-moderes, etc...). Dans une premiere partie, nous donnons une description detaillee des deux reacteurs. Dans la deuxieme partie, relative aux mesures de laplaciens, nous decrivons d'abord les mesures absolues de laplaciens (cartes de flux), puis les mesures relatives effectuees par la methode originale de remplacement progressif. Les resultats experimentaux sont rassembles dans le tableau VI. Dans la troisieme partie, nous rappelons un certain nombre d'autres mesures effectuees sur G1. (auteur)

  10. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  11. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  12. RB research reactor safety report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

  13. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

    The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

  14. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  15. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  17. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  18. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  19. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  20. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  1. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  2. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  4. Pb migration in the OKLO uranium deposit

    International Nuclear Information System (INIS)

    Gancarz, A.J.; Curtis, D.B.

    1979-01-01

    U-Pb and Pb isotopic data are presented which indicate that Pb is lost from host uraninite by diffusion, and that not only in situ uranogenic Pb but also the initial Pb is lost by diffusion. The conglomerate underlying the U deposit contains excess Pb and is both a transport zone and the repository for the Pb. 2 figures

  5. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

    International Nuclear Information System (INIS)

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

  6. Technology of nuclear reactors

    International Nuclear Information System (INIS)

    Ravelet, F.

    2016-01-01

    This academic report for graduation in engineering first presents operation principles of a nuclear reactor core. It presents core components, atomic nuclei, the notions of transmutation and radioactivity, quantities used to characterize ionizing radiations, the nuclear fission, statistical aspects of fission and differences between fast and slow neutrons, a comparison between various heat transfer fluids, the uranium enrichment process, and different types of reactor (boiling water, natural uranium and heavy water, pressurized water, and fourth generation). Then, after having recalled the French installed power, the author proposes an analysis of a typical 900 MWe nuclear power plant: primary circuit, reactor, fuel, spent fuel, pressurizer and primary pump, secondary circuit, aspects related to control-command, regulation, safety and exploitation. The last part proposes a modelling of the thermodynamic cycle of a pressurized water plant by using an equivalent Carnot cycle, a Rankine cycle, and a two-phase expansion cycle with drying-overheating

  7. Neutron source for a reactor

    International Nuclear Information System (INIS)

    Kobayashi, Hiromasa.

    1975-01-01

    Object: To easily increase a start-up power of a reactor without irradiation in other reactors. Structure: A neutron source comprises Cf 252 , a natural antimony rod, a layer of beryllium, and a vessel of neutron source. On upper and lower portion of Cf 252 are arranged natural antimony rods, which are surrounded by the Be layer, the entirety being charged into the vessel. The Cf 252 may emit neutron, has a half life more than a period of operating cycle of the reactor and is less deteriorated even irradiated by radioactive rays while being left within the reactor. The natural antimony rod is radioactivated by neutron from Cf 252 and neutron as reactor power increases to emit γ rays. The Be absorbs γ rays to emit the neutron. The antimony rod is irradiated within the reactor. Further, since the Cf 252 is small in neutron absorption cross section, it is hard to be deteriorated even while being inserted within the reactor. (Kamimura, M.)

  8. Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961); Pile AQUILON. Durcissement du spectre des neutrons dans les barreaux d'uranium et calcul des fissions epithermiques (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Durand -Smet, R; Lourme, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    - Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott '{beta}' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors) [French] - Un certain nombre de mesures effectuees dans la pile Aquilon ont permis d'etablir la distribution fine des flux thermique et epithermique dans les barreaux d'uranium, et d'en deduire les variations du spectre des neutrons dans l'uranium, le terme {beta} du spectre de Wescott moyen dans le barreau et le nombre de fissions epithermiques. En annexe, il est propose une definition nouvelle du coefficient de multiplication infini, qui fait intervenir les parametres epithermiques. (auteurs)

  9. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  10. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  12. Anaerobic treatment of palm oil mill effluent in batch reactor with digested biodiesel waste as starter and natural zeolite for microbial immobilization

    Science.gov (United States)

    Setyowati, Paulina Adina Hari; Halim, Lenny; Mellyanawaty, Melly; Sudibyo, Hanifrahmawan; Budhijanto, Wiratni

    2017-05-01

    Palm oil mill effluent (POME) is the wastewater discharged from sludge separation, sterilization, and clarification process of palm oil industries. Each ton of palm oil produces about half ton of high organic load wastewater. Up to now, POME treatment is done in lagoon, leaving major problems in land requirement and greenhouse gasses release. The increasing of palm oil production provokes the urgency of appropriate technology application in treating POME to prevent the greenhouse gasses emission while exploit POME as renewable energy source. The purposes of this study were firstly to test the effectiveness of using the digested biodiesel waste as the inoculum and secondly to evaluate the effectiveness of natural zeolite addition in minimizing the inhibitory effect in digesting POME. It was expected that the oil-degrading bacteria in the inoculum would shorten the adaptation period in digesting POME. Furthermore, the consortium formation of anaerobic bacteria accelerated by natural zeolite powder addition would increase the microbial resistance to the inhibitors contained in the POME. The batch digesters, containing 0 (control); 17; 38; and 63 g natural zeolite/g sCOD substrate were observed for 43 days. The result showed that zeolite addition did not give significant effect on sCOD reduction (97.3-98.6% of initial sCOD). Moreover, addition of immobilization media up to 17 g natural zeolite/g stimulated the acidification and biogas production up to 10% higher than control. The purity of methane produced with various amount of immobilization media did not differ for each variation, i.e. 50-54% v/v methane. The increasing amount of natural zeolite up to 63 g/g sCOD did not significantly enhance biogas product rate nor methane content.

  13. Effectiveness of solar disinfection using batch reactors with non-imaging aluminium reflectors under real conditions: Natural well-water and solar light.

    Science.gov (United States)

    Navntoft, C; Ubomba-Jaswa, E; McGuigan, K G; Fernández-Ibáñez, P

    2008-12-11

    Inactivation kinetics are reported for suspensions of Escherichia coli in well-water using compound parabolic collector (CPC) mirrors to enhance the efficiency of solar disinfection (SODIS) for batch reactors under real, solar radiation (cloudy and cloudless) conditions. On clear days, the system with CPC reflectors achieved complete inactivation (more than 5-log unit reduction in bacterial population to below the detection limit of 4CFU/mL) one hour sooner than the system fitted with no CPC. On cloudy days, only systems fitted with CPCs achieved complete inactivation. Degradation of the mirrors under field conditions was also evaluated. The reflectivity of CPC systems that had been in use outdoors for at least 3 years deteriorated in a non-homogeneous fashion. Reflectivity values for these older systems were found to vary between 27% and 72% compared to uniform values of 87% for new CPC systems. The use of CPC has been proven to be a good technological enhancement to inactivate bacteria under real conditions in clear and cloudy days. A comparison between enhancing optics and thermal effect is also discussed.

  14. Fusion reactor fuel processing

    International Nuclear Information System (INIS)

    Johnson, E.F.

    1972-06-01

    For thermonuclear power reactors based on the continuous fusion of deuterium and tritium the principal fuel processing problems occur in maintaining desired compositions in the primary fuel cycled through the reactor, in the recovery of tritium bred in the blanket surrounding the reactor, and in the prevention of tritium loss to the environment. Since all fuel recycled through the reactor must be cooled to cryogenic conditions for reinjection into the reactor, cryogenic fractional distillation is a likely process for controlling the primary fuel stream composition. Another practical possibility is the permeation of the hydrogen isotopes through thin metal membranes. The removal of tritium from the ash discharged from the power system would be accomplished by chemical procedures to assure physiologically safe concentration levels. The recovery process for tritium from the breeder blanket depends on the nature of the blanket fluids. For molten lithium the only practicable possibility appears to be permeation from the liquid phase. For molten salts the process would involve stripping with inert gas followed by chemical recovery. In either case extremely low concentrations of tritium in the melts would be desirable to maintain low tritium inventories, and to minimize escape of tritium through unwanted permeation, and to avoid embrittlement of metal walls. 21 refs

  15. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  16. Simultaneous E. coli inactivation and NOM degradation in river water via photo-Fenton process at natural pH in solar CPC reactor. A new way for enhancing solar disinfection of natural water.

    Science.gov (United States)

    Moncayo-Lasso, Alejandro; Sanabria, Janeth; Pulgarin, César; Benítez, Norberto

    2009-09-01

    Bacteria inactivation and natural organic matter oxidation in river water was simultaneously conducted via photo-Fenton reaction at "natural" pH ( approximately 6.5) containing 0.6 mg L(-1) of Fe(3+) and 10 mg L(-1) of H(2)O(2). The experiments were carried out by using a solar compound parabolic collector on river water previously filtered by a slow sand filtration system and voluntarily spiked with Escherichia coli. Fifty five percent of 5.3 mg L(-1) of dissolved organic carbon was mineralized whereas total disinfection was observed without re-growth after 24h in the dark.

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  19. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  20. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  1. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  2. Constant system for by-channel thermal-hydraulic calculation of fuel assembly operational conditions in reactors with natural and mixed convection

    International Nuclear Information System (INIS)

    Bogatyrev, I.L.; Bogoslovskaya, G.P.; Zhukov, A.V.; Sorokin, A.P.; Titov, P.A.

    1992-01-01

    System of constants for mass, impulse and energy conservation equations (drag, mixing, heat transfer coefficients, azimuthal unquality of temperature) is reported in region with small Re number for wide range of geometrical assembly parameters. This system can be used in subchannel calculations of assemblies with natural and mixed convection under conditions with loss of flow accident. The formulae are compared with experimental data. 30 refs.; 12 figs.; 1 tab

  3. Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for the Advanced Integral Type Reactor

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki; Lee, Un-Chul

    2004-01-01

    Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable the natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with the operation of the PRHRS. (authors)

  4. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

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

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  6. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  7. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  8. Reactor cooling apparatus

    International Nuclear Information System (INIS)

    Ogura, Kenji.

    1983-01-01

    Purpose: To increase natural convection flowrate in the reactor core upon interruption of a recycling pump by remarkably decreasing the flow resistance. Constitution: By-pass lines are disposed to a recycling pump in a primary coolant system and a second recycling pump in a secondary coolant system respectively, and a check valve and an isolation valve are attached to each of them. Each of the isolation valves is closed during normal operation and automatically opened when the number of rotation for each of the recycling pumps goes lower than a predetermined value. This can significantly decrease the flow resistance in the primary and secondary coolant systems upon interruption of the recycling pumps due to the entire loss of AC power source or the like to thereby increase the natural convection flowrate in the reactor core. (Sekiya, K.)

  9. Report of the Federal Ministry for the Environment, Nature Conservation, Buildings and Nuclear Safety (BMUB) on the topical peer review aging management in nuclear power plants and research reactors

    International Nuclear Information System (INIS)

    2017-01-01

    The report of the Federal Environmental Ministry (BMUB) on the topical peer review aging management in nuclear power plants and research reactors covers the following issues: comprehensive requirements for aging management and its implementation, electric cables, non accessible pipes, reactor pressure vessel, calandria/pressure tubes (CANDU), concrete containment, pre-stressed concrete reactor pressure vessel (AGR).

  10. TU electric reactor model verification

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1989-01-01

    Power reactor benchmark calculations using the code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles include gadolinia as a burnable absorber, natural uranium axial blankets, and increased water-to-fuel ratio. The calculated results for both low-power physics tests (boron end points, control rod worths, and isothermal temperature coefficients) and full-power operation (power distributions and boron letdown) are compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important physics parameters for power reactors

  11. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  12. Behaviour of palladium(II), platinum(IV), and rhodium(III) in artificial and natural waters: Influence of reactor surface and geochemistry on metal recovery

    Energy Technology Data Exchange (ETDEWEB)

    Cobelo-Garcia, Antonio [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)]. E-mail: antonio.cobelo-garcia@plymouth.ac.uk; Turner, Andrew [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Millward, Geoffrey E. [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom); Couceiro, Fay [School of Earth, Ocean and Environmental Sciences, University of Plymouth, Drake Circus, Plymouth PL4 8AA (United Kingdom)

    2007-03-07

    The recovery of dissolved platinum group elements (PGE: Pd(II), Pt(IV) and Rh(III)) added to Milli-Q[reg] water, artificial freshwater and seawater and filtered natural waters has been studied, as a function of pH and PGE concentration, in containers of varying synthetic composition. The least adsorptive and/or precipitative loss was obtained for borosilicate glass under most of the conditions employed, whereas the greatest loss was obtained for low-density polyethylene. Of the polymeric materials tested, the adsorptive and/or precipitative loss of PGE was lowest for fluorinated ethylene propylene (Teflon[reg]). The loss of Pd(II) in freshwater was significant due to its affinity for surface adsorption and its relatively low solubility. The presence of natural dissolved organic matter increases the recovery of Pd(II) but enhances the loss of Pt(IV). The loss of Rh(III) in seawater was significant and was mainly due to precipitation, whereas Pd(II) recovery was enhanced, compared to freshwater, because of its complexation with chloride. The results have important implications regarding protocols employed for sample preservation and controlled laboratory experiments used in the study of the speciation and biogeochemical behaviour of PGE.

  13. Establishing a Radiation Protection Programme for a Research Reactor

    International Nuclear Information System (INIS)

    Abdallah, M. M.

    2014-04-01

    The nature and intensity of radiation from the operation of a research reactor depend on the type of reactor, its design features and its operational history. The protection of workers from the harmful effect of radiation must therefore be of paramount importance to any operating organization of a research reactor. This project report attempts to establish an operational radiation protection programme for a research reactor using the Ghana Research Reactor-1 as a case study. (au)

  14. Astrid (fast breeder nuclear reactor)

    International Nuclear Information System (INIS)

    2014-01-01

    This document presents ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a French project of sodium-cooled fast breeder reactor, fourth generation reactor which should be fuelled by uranium 238 rather than uranium 235, and should therefore need less extracted natural uranium to produce electricity. The operation principle of fast breeder reactors is described. They notably directly consume plutonium, allow an easier radioactive waste management as they transform long life radioactive elements into shorter life elements by transmutation. The regeneration process is briefly described, and the various operation modes are evoked (iso-generator, sub-generator, and breeder). Some peculiarities of sodium-cooled reactors are outlined. The Astrid operation principle is described, its main design innovations outlined. Various challenges are discussed regarding safety of supply and waste processing, and the safety of future reactors. Major actors are indicated: CEA, Areva, EDF, SEIV Alcen, Toshiba, Rolls Royce, and Comex. Some key data are indicated: expected lifetime, expected availability rate, cost. The projected site is Marcoule and fast breeder reactors operated or under construction in the world are indicated. The document also proposes an overview of the background and evolution of reactors of 4. generation

  15. Research reactor`s role in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1996-12-31

    After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

  16. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  18. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  20. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  1. Nuclear Reactor Sharing Program

    International Nuclear Information System (INIS)

    1994-01-01

    The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

  2. Advanced converters and reactors

    International Nuclear Information System (INIS)

    Haefele, W.; Kessler, G.

    1984-01-01

    As Western Europe and most countries of the Asia-Pacific region (except Australia) have only small natural uranium resources, they must import nuclear fuel from the major uranium supplier countries. The introduction of advanced converter and breeder reactor technology allows a fuel utilization of a factor of 4 to 100 higher than with present low converters (LWRs) and will make uranium-importing countries less vulnerable to price jumps and supply stops in the uranium market. In addition, breeder-reactor technology will open up a potential that can cover world energy requirements for several thousand years. The enormous development costs of advanced converter and breeder technologies can probably be raised only by highly industrialized countries. Those highly industrialized countries that have little or no uranium resources (Western Europe, Japan) will probably be the first to introduce this advanced reactor technology on a commercial scale. A number of small countries and islands will need only small power reactors with inherent safety capabilities, especially in the beginning of their nuclear energy programs. For economic reasons, the fuel cycle services should come from large reprocessing centers of countries having sufficiently large nuclear power programs or from international fuel cycle centers. (author)

  3. Fuel recycling and 4. generation reactors

    International Nuclear Information System (INIS)

    Devezeaux de Lavergne, J.G.; Gauche, F.; Mathonniere, G.

    2012-01-01

    The 4. generation reactors meet the demand for sustainability of nuclear power through the saving of the natural resources, the minimization of the volume of wastes, a high safety standard and a high reliability. In the framework of the GIF (Generation 4. International Forum) France has decided to study the sodium-cooled fast reactor. Fast reactors have the capacity to recycle plutonium efficiently and to burn actinides. The long history of reprocessing-recycling of spent fuels in France is an asset. A prototype reactor named ASTRID could be entered into operation in 2020. This article presents the research program on the sodium-cooled fast reactor, gives the status of the ASTRID project and present the scenario of the progressive implementation of 4. generation reactors in the French reactor fleet. (A.C.)

  4. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  5. Research reactors; Les piles de recherche

    Energy Technology Data Exchange (ETDEWEB)

    Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

    1955-07-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  6. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  7. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  8. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  10. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  12. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  13. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  14. Fuel assembly for FBR type reactor and reactor core thereof

    International Nuclear Information System (INIS)

    Kobayashi, Kaoru.

    1998-01-01

    The present invention provides a fuel assembly to be loaded to a reactor core of a large sized FBR type reactor, in which a coolant density coefficient can be reduced without causing power peaking in the peripheral region of neutron moderators loaded in the reactor core. Namely, the fuel assembly for the FBR type reactor comprises a plurality of fission product-loaded fuel rods and a plurality of fertile material-loaded fuel rods and one or more rods loading neutron moderators. In this case, the plurality of fertile material-loaded fuel rods are disposed to the peripheral region of the neutron moderator-loaded rods. The plurality of fission product-loaded fuel rods are disposed surrounding the peripheral region of the plurality of fertile material-loaded fuel rods. The neutron moderator comprises zirconium hydride, yttrium hydride and calcium hydride. The fission products are mixed oxide fuels. The fertile material comprises depleted uranium or natural uranium. (I.S.)

  15. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  16. Technique of nuclear reactors controls

    International Nuclear Information System (INIS)

    Weill, J.

    1953-12-01

    This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [fr

  17. Heat extraction from HTGR reactor

    International Nuclear Information System (INIS)

    Balajka, J.; Princova, H.

    1986-01-01

    The analysis of an HTGR reactor energy balance showed that steam reforming of natural gas or methane is the most suitable process of utilizing the high-temperature heat. Basic mathematical relations are derived allowing to perform a general energy balance of the link between steam reforming and reactor heat output. The results of the calculation show that the efficiency of the entire reactor system increases with increasing proportion of heat output for steam reforming as against heat output for the steam generator. This proportion, however, is limited with the output helium temperature from steam reforming. It is thus always necessary to use part of the reactor heat output for the steam cycle involving electric power generation or low-potential heat generation. (Z.M.)

  18. Economic analysis of EBT reactor

    International Nuclear Information System (INIS)

    Woo, J.T.; Uckan, N.A.; Lidsky, L.M.

    1977-01-01

    In order to establish the economic potential of the Elmo Bumpy Torus (EBT) reactor, two independent system-costing models have been developed. Both models predict capital costs of approximately $400/kW(th). These relatively low costs reflect the simplicity of the EBTR design. In particular, the modular nature of the individual blanket-shield segments, the low costs ''accelerator style'' containment building, high beta, and steady-state operation lead to relatively low reactor costs. A detailed cost breakdown for subsystems is analyzed. High cost and high uncertainty subsystems are identified to direct further emphasis into those areas. The calculated capital costs for the EBT reactor are compared with those costs quoted for tokamak reactors

  19. Thermosyphoning in the CANDU reactor

    International Nuclear Information System (INIS)

    Spinks, N.J.; Wright, A.C.D.; Caplan, M.Z.; Prawirosoehardjo, S.; Gulshani, P.

    1984-01-01

    Thermosyphoning is defined as the natural convective flow of primary coolant over the boilers. It is the predicted mode of heat transport from core to boilers in many postulated scenarios for CANDU reactor safety analysis. The scenarios encompass a wide range of boundary conditions in reactor power, secondary temperature and primary coolant inventory. Loss of pumping of the primary coolant is a common feature. Thermosyphoning is single or two-phase depending on the boundary conditions. The paper describes the important thermohydraulic characteristics of thermosyphoning in CANDU reactors with emphasis on two-phase thermosyphoning. It utilizes predictions of a transient thermohydraulics computer code and describes experiments done for the purpose of verifying these predictions. Predictions are compared with single-phase thermosyphoning tests done during commissioning of the Gentilly-2 and Point Lepreau CANDU 600 reactors. (orig.)

  20. Basic natural circulation characteristics of SBWR

    International Nuclear Information System (INIS)

    Kuran, S.; Soekmen, C. N.

    2001-01-01

    Natural circulation is an important passive heat removal mechanism for both existing and next generation light water reactors. Simplified Boiling Water Reactor (SBWR) is one of the advanced light water reactors that rely on natural circulation for normal as well as emergency core cooling. In this study, basic natural circulation characteristics of this reactor are examined on a flow loop that simulates the operation of SBWR. On this model, effect of system operating parameters on the steady state natural circulation characteristics inside the loop is studied via solving the transcendental equation for loop flow rate

  1. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  4. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  5. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  6. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  7. Nuclear Power: Outlook for New U.S. Reactors

    National Research Council Canada - National Science Library

    Parker, Larry; Holt, Mark

    2007-01-01

    .... The renewed interest in nuclear power has resulted primarily from higher prices for natural gas, improved operation of existing reactors, and uncertainty about future restrictions on coal emissions...

  8. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  11. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  13. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  15. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  16. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  17. SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R. A.

    1963-10-15

    The nature of nuclear power reactors demands an exceptionally high degree of seismic integrity. Considerations involved in defining earthquake resistance requirements are discussed. Examples of seismic design criteria and applications of the spectrum technique are described. (auth)

  18. General aspects of CAREM-25 reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

    2000-01-01

    CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

  19. General Aspects of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

    2000-01-01

    CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

  20. Simulation of a pool type research reactor

    International Nuclear Information System (INIS)

    Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes

    2011-01-01

    Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)

  1. Dissolution studies of natural analogues spent fuel and U(VI)-Silicon phases of and oxidative alteration process

    International Nuclear Information System (INIS)

    Perez Morales, I.

    2000-01-01

    In order to understand the long-term behavior of the nuclear spent fuel in geological repository conditions, we have performed dissolution studies with natural analogues to UO 2 as well as with solid phases representatives of the oxidative alteration pathway of uranium dioxide, as observed in both natural environment and laboratory studies. In all cases, we have studied the influence of the bicarbonate concentration in the dissolution process, as a first approximation to the groundwater composition of a granitic environment, where carbonate is one of the most important complexing agents. As a natural analogue to the nuclear spent fuel some uraninite samples from the Oklo are deposit in Gabon, where chain fission reactions took place 2000 millions years ago, as well as a pitchblende sample from the mine Fe ore deposit, in Salamanca (spain) have been studied. The studies have been performed at 25 and 60 deg C and 60 deg C, and they have focussed on the determination of both the thermodynamic and the kinetic properties of the different samples studied, using batch and continuous experimental methodologies, respectively. (Author)

  2. TRIGA reactor operating experience

    International Nuclear Information System (INIS)

    Anderson, T.V.

    1970-01-01

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  3. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  4. Southern Africa - a giant natural photochemical reactor

    CSIR Research Space (South Africa)

    Diab, RD

    2006-04-01

    Full Text Available photochemical reactor’ are abundant sources of ozone precursors (biomass burning, lightning, biogenic and urban-industrial sources), and meteorological conditions that promote anticyclonic recirculation on a subhemispheric scale....

  5. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  6. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

    Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions, thus preventing further complications by alerting opeators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Dr. Thie hopes to further, through detailed explanations and over 70 illustrations, the acceptance of the use of noise analysis by the nuclear utility industry. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussions of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

  7. Status of advanced technologies for CANDU reactors

    International Nuclear Information System (INIS)

    Lipsett, J.J.

    1989-01-01

    The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

  8. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    1965-01-01

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  9. CANDU reactor - supporting the nuclear renaissance

    International Nuclear Information System (INIS)

    Oberth, R.

    2010-01-01

    'Full text:' The CANDU reactor has proven to be a strong performer in both the Canada, with 22 units constructed in Ontario, New Brunswick and Quebec, as well as in Argentina, Korea, Romania and China where a further nine units are operating and two in the planning stage. The average lifetime capacity factor of the CANDU reactor fleet is 89%. The last seven CANDU projects in Korea, China, and Romania have been completed on budget and on schedule. CANDU reactors have the highest uranium utilization efficiency measures as electricity output per ton of uranium mined. The CANDU fuel channel design using on-power fuelling and a heavy water moderator enables flexible fueling options - from the current natural uranium option to burning uranium recovered from used LWR reactor fuel and even a thorium-based fuel. AECL and the CANDU reactor are poised to participate in the worldwide construction at least 250 new reactors over the next 20 years. (author)

  10. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    Hatcher, S.R.; McDonnell, F.N.; Griffiths, J.; Boczar, P.G.

    1987-01-01

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  11. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  12. The CANDUR Reactor - The Practical Path to RU and TH use in Nuclear Reactors

    International Nuclear Information System (INIS)

    Kuran, Sermet; Yang, Dezi

    2012-01-01

    The CANDU heavy water reactor has unrivalled flexibility for using a variety of fuels, such as Natural Uranium (NU), Low Enriched Uranium (LEU), Recycled Uranium (RU), Mixed Oxide (MOX), and Thorium (Th). Recently, this unique CANDU reactor feature attracted considerable attention due to favourable commercial, environmental and strategic needs. This paper summarizes the solid progress over the last three years and outlines CANDU Energy Incorporated's (CEI) multi-stage vision of utilizing various fuels in currently operational and new build CANDU reactors. In CEI's fuel-cycle vision, CANDU reactors will operate in conjunction with other reactor types and use advanced fuels to produce more energy and ensure the most efficient and least costly method of utilizing Light Water Reactor (LWR) used fuel. With this vision and the tandem goal of systematic adoption of Thorium based fuels, CANDU reactors will be a strong technology partner in ensuring the availability of long-term stable resources for nuclear power plants

  13. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  14. Reactor accidents and the environment

    International Nuclear Information System (INIS)

    Beattie, J.R.; Griffiths, R.F.; Kaiser, G.D.; Kinchin, G.H.

    1978-01-01

    This is a condensed version of a paper, entitled 'The Environmental Impact of Radioactive Releases from Accidents in Nuclear Power Reactors', by the authors, presented to the Nuclear Energy Panel of the International Atomic Energy Agency/United Nations Environmental Programme. Headings include - Effects of ionising radiation on man; number of deaths expected from leukaemia and other cancers; risk estimates for incidence of benign nodules and thyroid cancer; maximum permissible levels and emergency levels of radiation and radioactivity; ICRP recommended dose limits for members of the general public; atmospheric dispersion and modelling; ICRP emergency reference levels for 1 131 , Cs 137 , Ru 106 and Sr 90 ; environmental consequences of accidental releases from nuclear power reactors; environmental impact of accidents to Magnox gas-cooled reactors; environmental impact of accidents to advanced gas-cooled reactors; environmental impact of accidents to fast reactors; and nature of risks. consequences are examined in terms of early and late biological effects on man, and contamination of land areas. Serious accidents are of low probability of occurrence, and the risk of accidents to nuclear power reactors is estimated to be very small. 43 references. (U.K.)

  15. An evaluation of the dissolution process of natural uranium ore as an analogue of nuclear fuel

    International Nuclear Information System (INIS)

    Stern, V.H.

    1991-08-01

    The assumption of congruent dissolution of uraninite as a mechanism for the dissolution behaviour of spent fuel was critically examined with regard to the fate of toxic radionuclides. The fission and daughter products of uranium are typically present in spent unreprocessed fuel rods in trace abundances. The principles of trace element geochemistry were applied in assessing the behaviour of these radionuclides during fluid/solid interactions. It is shown that the behaviour of radionuclides in trace abundances that reside in the crystal structure can be better predicted from the ionic properties of these nuclides rather than from assuming that they are controlled by the dissolution of uraninite. Geochemical evidence from natural uranium ore deposits (Athabasca Basin, Northern Territories of Australia, Oklo) suggests that in most cases the toxic radionuclides are released from uraninite in amounts that are independent of the solution behaviour of uranium oxide. Only those elements that have ionic and thus chemical properties similar to U 4+ , such as plutonium, americium, cadmium, neptunium and thorium can be satisfactorily modelled by the solution properties of uranium dioxide and then only if the environment is reducing. (84 refs., 7 tabs.)

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  19. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  20. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  2. Argentinean integrated small reactor design and scale economy analysis of integrated reactor

    International Nuclear Information System (INIS)

    Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

    2000-01-01

    This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

  3. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  4. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  5. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  6. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  7. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  8. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  9. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  10. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  11. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  12. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  13. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  14. Solid State Reactor Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mays, G.T.

    2004-03-10

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas

  15. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

  16. Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

  17. Reactor safety

    International Nuclear Information System (INIS)

    Meneley, D.A.

    The people of Ontario have begun to receive the benefits of a low cost, assured supply of electrical energy from CANDU nuclear stations. This indigenous energy source also has excellent safety characteristics. Safety has been one of the central themes of the CANDU development program from its very beginning. A great deal of work has been done to establish that public risks are small. However, safety design criteria are now undergoing extensive review, with a real prospect of more stringent requirements being applied in the future. Considering the newness of the technology it is not surprising that a consensus does not yet exist; this makes it imperative to discuss the issues. It is time to examine the policies and practice of reactor safety management in Canada to decide whether or not further restrictions are justified in the light of current knowledge

  18. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  20. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)