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Sample records for piqua organic moderated reactor

  1. 2016 Annual Inspection and Radiological Survey Results for the Piqua, Ohio, Decommissioned Reactor Site, July 2016

    Energy Technology Data Exchange (ETDEWEB)

    Zimmerman, Brian [USDOE Office of Legacy Management, Washington, DC (United States); Miller, Michele [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-07-01

    This report presents the findings of the annual inspection and radiological survey of the Piqua, Ohio, Decommissioned Reactor Site (site). The decommissioned nuclear power demonstration facility was inspected and surveyed on April 15, 2016. The site, located on the east bank of the Great Miami River in Piqua, Ohio, was in fair physical condition. There is no requirement for a follow-up inspection, partly because City of Piqua (City) personnel participated in a March 2016 meeting to address reoccurring safety concerns. Radiological survey results from 104 locations revealed no removable contamination. One direct beta activity reading in a floor drain on the 56-foot level (1674 disintegrations per minute [dpm]/100 square centimeters [cm2]) exceeded the minimum detectable activity (MDA). Beta activity has been detected in the past at this floor drain. The reading was well below the action level of 5000 dpm/100 cm2.

  2. Post-Construction Testing of the Elk River, Hallam and Piqua Power Reactor Plants; Essais apres construction des centrales nucleaires d'Elk River, de Hallam et de Piqua; Predehkspluatatsionnoe ispytanie Ehlk-riverskoj, Khehlpemskoj i Pikuaskoj ehnergeticheskikh reaktornykh ustanovok; Ensayos posteriores a la construccion de las centrales nucleoelectricas de Elk River, Hallam y Piqua

    Energy Technology Data Exchange (ETDEWEB)

    Pursel, C. A. [United States Atomic Energy Commission, Argonne, IL (United States)

    1963-10-15

    Actual experience gained during the post-construction testing of three nuclear power plants, under the USAEC Power Reactor Demonstration Program, may permit some generalizations concerning this phase of plant construction and operation. The three plants, Elk River Reactor (ERR), Hallam Nuclear Power Facility (HNPF), and the Piqua Nuclear Power Facility (PNPF), represent three different reactor concepts: natural-circulation boiling water, sodiumgraphite, and organic cooled and moderated, respectively. The post-construction testing period included the time between the end of construction (erection of structures and installation of equipment) and the beginning of power operation (generation of significant net electrical power). The tests were intended to: (a) verify the performance characteristics of the as-installed equipment; (b) obtain initial criticality and reactivity coefficient measurements; and (c) determine reactor physics and plant performance characteristics at a sequence of increasing power levels. .The experience gained can be reported in six separate but interrelated categories: (1) schedule; (2) costs; (3) staffing requirements; (4) procedures; (5) equipment performance (including malfunctions); and (6) actual, as compared to predicted, system performance characteristics. The average project staffing, including craftsmen, operators, supervisors, technical support and trainees, was approximately 50 for ERR, 115 for HNPF, and 60 for PNPF. Detailed written Pre-operational Test Procedures were prepared for each major component and system. To the maximum possible extent, all tests were performed before fuel loading and operation of the integrated plant. Authorization procedures (duplicates of the licensing procedures for non-USAEC-owned plants) were in progress during almost all of the post-construction testing periods. The time required for post-construction testing of each of these plants significantly exceeded the original estimates. The tests disclosed

  3. District heating and cooling systems for communities through power plant retrofit and distribution network, city of Piqua, Ohio

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-01-01

    The feasibility analysis and evaluation of the Piqua, Ohio District Heating and Cooling Demonstration program is being conducted by the Piqua Municipal Power Co., the Piqua Law Dept., the Public Works Dept., a firm of economic analysts, and the Georgia Tech Engineering Dept. This volume contains information on the organization and composition of the demonstration team; characterization of the Piqua community; and the technical, environmental, institutional; financial, and economic assessments of the project. (LCL)

  4. Annual monitoring and surveillance report for Piqua Nuclear Power Facility, Piqua, Ohio

    International Nuclear Information System (INIS)

    Mosho, G.D.

    1991-12-01

    This report discusses the decommissioned Piqua Nuclear Power Facility which is located in Piqua, Ohio near the Greater Miami River. The Facility was built by the US Atomic Energy Commission (now U. S. Department of Energy) and was operated from 1963 to 1966. The reactor was retired prior to 1970 and the facility was leased to the city of Piqua for use as offices and equipment storage. In December 1991, a radiological survey was done of the facility to document its radiological condition. The data show that all radiological parameters measured were essentially the same as that found in the natural environment. The only exception was that low levels of radioactive contamination were detected in one drain on the 56.5 ft elevation, but the radiation exposure rate in that area was also typical of natural background

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

    International Nuclear Information System (INIS)

    Hine, R.E.

    1980-09-01

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

  6. Moderator for nuclear reactor

    International Nuclear Information System (INIS)

    Milgram, M.S.; Dunn, J.T.; Hart, R.S.

    1995-01-01

    This invention relates to a moderator for a nuclear reactor and more specifically, to a composite moderator. A moderator is designed to slow down, or thermalize, neutrons which are released during nuclear reactions in the reactor fuel. Pure or almost pure materials like light water, heavy water, beryllium or graphite are used singly as moderators at present. All these materials, are used widely. Graphite has a good mechanical strength at high temperatures encountered in the nuclear core and therefore is used as both the moderator and core structural material. It also exhibits a low neutron-capture cross section and high neutron scattering cross section. However, graphite is susceptible to attach by carbon dioxide and/or oxygen where applicable, and releases stress energy under certain circumstances, although under normal operating conditions these reactions can be controlled. (author). 1 tab

  7. FLUID MODERATED REACTOR

    Science.gov (United States)

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

    1957-10-22

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

  8. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

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

    1957-09-17

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

  9. Program of chemical research organic moderators and coolants in regard with the project of the Spanish DON-Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez Cellini, R

    1961-07-01

    In order to select the most suitable mixtures for cooling the Reactor DON, in project, within the group of polyphenyl compounds, which at present time prevails in organic nuclear power reactor philosophy, the following subjects are going to be studied: 1. Static and dynamic experiences of thermal and radiolytic stability of commercial polyphenyl mixtures. 2. Studies on purification of the damaged coolant and search for possible applications of the residual polymers. 3. Analytical control and measurement of the physical constants of the polyphenyl mixtures. (Author)

  10. Program of chemical research on organic moderators and coolants in regard with the project of the Spanish DON-Reactor

    International Nuclear Information System (INIS)

    Fernandez Cellini, R.

    1961-01-01

    In order to select the most suitable mixtures for cooling the Reactor DON, in project, within the group of polyphenyl compounds, which at present time prevails in organic nuclear power reactor philosophy, the following subjects are going to be studied: 1. Static and dynamic experiences of thermal and radiolytic stability of commercial polyphenyl mixtures. 2. Studies on purification of the damaged coolant and search for possible applications of the residual polymers. 3. Analytical control and measurement of the physical constants of the polyphenyl mixtures. (Author)

  11. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

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

  12. Overview moderator material for nuclear reactor components

    International Nuclear Information System (INIS)

    Mairing Manutu Pongtuluran; Hendra Prihatnadi

    2009-01-01

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

  13. Moderator circulation in CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Hussein, M.A.

    1989-01-01

    A two-dimensional computer code that is capable of predicting the moderator flow and temperature distribution inside CANDU calandria is presented. The code uses a new approach to simulate the calandria tube matrix by blocking the cells containing the tubes in the finite difference mesh. A jet momentum-dominant flow pattern is predicted in the nonisothermal case, and the effect of the buoyancy force, resulting from nuclear heating, is found to enhance the speed of circulation. Hot spots are located in low-velocity areas at the top of the calandria and below the inlet jet level between the fuel channels. A parametric study is carried out to investigate the effect of moderator inlet velocity,moderator inlet nozzle location, and geometric scaling. The results indicate that decreasing the moderator inlet velocity has no significant influence on the general features of the flow pattern (i.e., momentum dominant); however, too many high-temperature hot spots appear within the fuel channels

  14. Method of operating heavy water moderated reactors

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1980-01-01

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

  15. Moderator heat recovery of CANDU reactors

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  16. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

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

    1964-01-01

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

  17. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

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

  18. High conversion heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyawaki, Yoshio; Wakabayashi, Toshio.

    1989-01-01

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

  19. Design of an organic simplified nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

    2016-08-15

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  20. Design of an Organic Simplified Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  1. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. Graphite moderated reactor for thermoelectric generation

    International Nuclear Information System (INIS)

    Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

    1998-01-01

    Fuel rods filled with cladded fuel particles distributed and filled are buried each at a predetermined distance in graphite blocks situated in a reactor core. Perforation channels for helium gas as coolants are formed to the periphery thereof passing through vertically. An alkali metal thermoelectric power generation module is disposed to the upper lid of a reactor container while being supported by a securing receptacle. Helium gas in the coolant channels in the graphite blocks in the reactor core absorbs nuclear reaction heat, to be heated to a high temperature, rises upwardly by the reduction of the specific gravity, and then flows into an upper space above the laminated graphite block layer. Then the gas collides against a ceiling and turns, and flows down in a circular gap around the circumference of the alkali metal thermoelectric generation module. In this case, it transfers heat to the alkali metal thermoelectric generation module. (I.N.)

  3. Solid methane cold moderator for the IBR-2 reactor

    International Nuclear Information System (INIS)

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

    1997-09-01

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

  4. Moderator behaviour and reactor internals integrity at Atucha I NPP

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Moderator behaviour and reactor internals integrity at Atucha I NPP

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  6. The high moderating ratio reactor using 100% MOX reloads

    International Nuclear Information System (INIS)

    Barbrault, P.

    1994-06-01

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

  7. Measuring device for the temperature coefficient of reactor moderators

    International Nuclear Information System (INIS)

    Nakano, Yuzo.

    1987-01-01

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

  8. UK methods for studying fuel management in water moderated reactors

    International Nuclear Information System (INIS)

    Fayers, F.J.

    1970-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

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

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Ishikawa, Nobuyuki; Iwamura, Takamichi

    2001-09-01

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

  11. Kinetics of Pressurized Water Reactors with Hot or Cold Moderators

    Energy Technology Data Exchange (ETDEWEB)

    Norinder, O

    1960-11-15

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

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

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

    2000-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

  14. Reactivity margins in heavy water moderated production reactors

    International Nuclear Information System (INIS)

    Benton, F.D.

    1981-11-01

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

  15. 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)

  16. Physical particularities of nuclear reactors using heavy moderators of neutrons

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  17. Physical particularities of nuclear reactors using heavy moderators of neutrons

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  18. Fluid moderator control system reactor internals distribution system

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  19. Fissile fuel assembly for a sub-moderated nuclear reactor

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  20. Nuclear calculation methods for light water moderated reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1961-02-01

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

  1. Study on core design for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Okubo, Tsutomu

    2002-01-01

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

  2. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  3. Research on organic liquids as reactor coolants. Status report from Hungary

    International Nuclear Information System (INIS)

    Kiss, I.

    1967-01-01

    The organic-moderated and cooled nuclear reactor concept has stimulated extensive activities in numerous different areas of research. Investigations started in Hungary in 1958 do not cover all topics of interest in organic reactors and so far no projects have been started to build such a reactor. Since OMRE and other organic reactor experiments have already shown the potential use of organic materials as reactor coolants and moderators, efforts have been focused rather on the investigation and solution of certain specific particular problems and also on economic aspects. One of the most important objectives seems to be a better knowledge of the radiolytic heat transfer and neutron physics behaviour of organic liquids. In Hungary the following topics were selected for investigation: Radiation stability of organic compounds and their mixtures; Heat-transfer studies; Investigations on the moderating parameters of organic liquids; Economic analysis of the possible use of organic reactors for process heat application

  4. Research on organic liquids as reactor coolants. Status report from Hungary

    Energy Technology Data Exchange (ETDEWEB)

    Kiss, I [Central Research Institute for Physics, Budapest (Hungary)

    1967-01-01

    The organic-moderated and cooled nuclear reactor concept has stimulated extensive activities in numerous different areas of research. Investigations started in Hungary in 1958 do not cover all topics of interest in organic reactors and so far no projects have been started to build such a reactor. Since OMRE and other organic reactor experiments have already shown the potential use of organic materials as reactor coolants and moderators, efforts have been focused rather on the investigation and solution of certain specific particular problems and also on economic aspects. One of the most important objectives seems to be a better knowledge of the radiolytic heat transfer and neutron physics behaviour of organic liquids. In Hungary the following topics were selected for investigation: Radiation stability of organic compounds and their mixtures; Heat-transfer studies; Investigations on the moderating parameters of organic liquids; Economic analysis of the possible use of organic reactors for process heat application.

  5. The liquid hydrogen moderator at the NIST research reactor

    International Nuclear Information System (INIS)

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

    1997-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F

    1984-07-01

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

  7. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1984-01-01

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

  9. Structure optimization of CFB reactor for moderate temperature FGD

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  10. The thorium fuel cycle in water-moderated reactor systems

    International Nuclear Information System (INIS)

    Critoph, E.

    1977-01-01

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

  11. Power control device for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Matsushima, Hidesuke; Masuda, Hiroyuki.

    1978-01-01

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

  12. Progress in design study on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu; Kugo, Teruhiko; Shimada, Shoichiro; Shirakawa, Toshihisa; Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takeda, Renzo [Hitachi Ltd., Tokyo (Japan); Yokoyama, Tsugio [Toshiba Corp., Kawasaki, Kanagawa (Japan); Hibi, Koki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Wada, Shigeyuki [Japan Atomic Power Co., Tokyo (Japan)

    2000-06-01

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

  13. Core design study on reduced-moderation water reactors

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  14. Plutonium multi-recycling in increased moderating ratio reactors (IMR)

    International Nuclear Information System (INIS)

    Barbrault, P.; Larderet, P.

    1998-01-01

    The large core of the future jointly defined European PWR (EPR), would be compatible with an increased Moderating Ratio (MR) enabling better plutonium burnout. The purpose of current work on the subject is to assess plutonium multi-recycling possibilities in IMR reactors. What additional operating constraints would be involved under normal and accidental conditions and are they acceptable? The conclusion is that Plutonium multi-recycling in a PWR of the type envisaged for the EPR raises no major problems under the following conditions: use of an IMR MOX core, enhancing both plutonium burnout and absorber efficiency; use of enriched boron in both the primary coolant soluble boron and the B4C boron carbide in the control rods. Deeper investigation should be performed concerning the partial or total core drain-out, in view of the high total Pu concentrations involved (13%) and the types of core considered (100% MOX). (author)

  15. Summary report of the 7th reduced-moderation water reactor workshop

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Nabeshima, Kunihiko; Uchikawa, Sadao

    2005-08-01

    As a research on the future innovative water reactor, the development of Reduced-Moderation Water Reactors (RMWRs) has been performed in Japan Atomic Energy Research Institute (JAERI). The workshop on RMWRs is aiming at information exchange between JAERI and other organizations such as universities, laboratories, utilities and vendors, and has been held every year since 1998. The 7th workshop was held on March 5, 2004 under the joint auspices of JAERI and North Kanto branch of Atomic Energy Society of Japan. The program of the workshop was composed of 5 lectures and an overall discussion time. The workshop started with the lecture by JAERI on the status and future program of PMWR research and development, followed by the two presentations by JAERI and Japan Nuclear Cycle Development Institute, respectively, on the investigation and evaluation of water cooled reactor in Feasibility Study Program on Commercialized Fast Reactor Systems. The lectures were also made on the Japan's nuclear fuel cycle and scenarios for RMWRs deployment by JAERI, and on the next generation reactor development activity by Hitachi, Ltd. The main subjects of the overall discussion time were Na cooled fast reactor, deployment effects of RMWRs and the future plan of the RMWR research and development. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture, as well as of the discussion time. In addition in the Appendices, there are included presentation handouts of each lecture, program of the workshop and the participants list. (author)

  16. Research on Reduced-Moderation Water Reactor (RMWR)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Shimada, Shoichiro

    1999-11-01

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

  17. Research on Reduced-Moderation Water Reactor (RMWR)

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

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

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

    International Nuclear Information System (INIS)

    Kulikov, S; Shabalin, E

    2012-01-01

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

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

    Science.gov (United States)

    Kulikov, S.; Shabalin, E.

    2012-03-01

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

  20. Calculational assessment of critical experiments with mixed-oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.; Funabashi, H.

    1987-01-01

    Critical experiments have been conducted with organically moderated mixed-oxide (MOX) fuel pin assemblies at the Pacific Northwest Lab. Critical Mass Lab. These experiments are part of a joint exchange program between the US Dept. of Energy and the Power Reactor and Nuclear Fuel Development Corp. of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organically moderated systems. Calculations presented in this paper indicated that the Scale code system and the 27-energy-group cross-section library accurately compute k/sub eff/ for organically moderated MOX fuel pin assemblies. Furthermore, the reactivity of an organic solution with a 32 vol % TBP/68 vol% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water

  1. Calculational assessment of critical experiments with mixed oxide fuel pin arrays moderated by organic solution

    International Nuclear Information System (INIS)

    Smolen, G.R.

    1987-01-01

    Critical experiments have been conducted with organic-moderated mixed oxide (MOX) fuel pin assemblies at the Pacific Northwest Laboratory (PNL) Critical Mass Laboratory (CML). These experiments are part of a joint exchange program between the United States Department of Energy (USDOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan in the area of criticality data development. The purpose of these experiments is to benchmark computer codes and cross-section libraries and to assess the reactivity difference between systems moderated by water and those moderated by an organic solution. Past studies have indicated that some organic mixtures may be better moderators than water. This topic is of particular importance to the criticality safety of fuel processing plants where fissile material is dissolved in organic solutions during the solvent extraction process. In the past, it has been assumed that the codes and libraries benchmarked with water-moderated experiments were adequate when performing design and licensing studies of organic-moderated systems. Calculations presented in this paper indicated that the SCALE code system and the 27-energy-group cross-section accurately compute k-effectives for organic moderated MOX fuel-pin assemblies. Furthermore, the reactivity of an organic solution with a 32-vol-% TBP/68-vol-% NPH mixture in a heterogeneous configuration is the same, for practical purposes, as water. 5 refs

  2. 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)

  3. Studies on gadolinium precipitation in moderator system of nuclear reactor

    International Nuclear Information System (INIS)

    Joshi, Akhilesh C.; Rajesh, Puspalata; Rufus, A.L.; Velmurugan, S.

    2015-01-01

    Gadolinium is used in the moderator system of many Pressurised Heavy Water Reactors (PHWRs) for start-up, shut-down and reactivity control during operation. It is very much essential to maintain gadolinium concentration in the system as desired. It has been reported that gadolinium gets precipitated in as oxalate in carbonated water under the influence of γ-radiation. Hence, studies were carried out to investigate the effect of dose, presence of other metal ions and metal surfaces on the precipitation of gadolinium. The results showed that the amount of carboxylic acids viz., formic acid and oxalic acid, formed due to radiolysis is dependent on the dose and that the curve passes though a maxima. Gadolinium is added in higher concentration in Advanced Heavy Water Reactor. So, experiments with high concentration of gadolinium were also carried out. Ultra pure water saturated with high purity CO 2 containing gadolinium and desired ion/surface was irradiated with γ-radiation from 60 Co source at 25°C to doses ranging from 2.5-16.6 Mrad. At lower doses, formation of carboxylic acids takes place but as the dose increases, decomposition of these acids starts and hence the concentration Vs dose passes through a maximum. It was found that precipitation of gadolinium as oxalate occurred at lower doses. At higher doses, it was seen that pH of the solution decreases and hence solubility of gadolinium oxalate increases. It was also observed that the amount of gadolinium precipitated varied linearly with the initial concentration of gadolinium varying from 2 ppm to 20 ppm. While for gadolinium concentration from 20 ppm to 400 ppm, gadolinium in particulate form was observed. The amount of carboxylic acids formed depends on the nature of cations present in solution. It was found that the amount of oxalic acid formed in the case of gadolinium was more than that formed in the case of sodium. Presence of metal oxides such as ZrO 2 formed over zircoloy surfaces was found to

  4. Status of research and development on reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Iwamura, Takamichi

    2002-01-01

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

  5. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  10. 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

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  12. Study on the effect of moderator density reactivity for Kartini reactor

    International Nuclear Information System (INIS)

    Budi Rohman; Widarto

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-06-01

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

  18. Moderator clean-up system in a heavy water reactor

    International Nuclear Information System (INIS)

    Sasada, Yasuhiro; Hamamura, Kenji.

    1983-01-01

    Purpose: To decrease the fluctuation of the poison concentration in heavy water moderator due to a heavy water clean-up system. Constitution: To a calandria tank filled with heavy water as poison-containing moderators, are connected both end of a pipeway through which heavy water flows and to which a clean-up device is provided. Strongly basic resin is filled within the clean-up device and a cooler is disposed to a pipeway at the upstream of the clean-up device. In this structure, the temperature of heavy water at the inlet of the clean-up device at a constant level between the temperature at the exit of the cooler and the lowest temperature for the moderator to thereby decrease the fluctuation in the poison concentration in the heavy water moderator due to the heavy water clean-up device. (Moriyama, K.)

  19. Calculation of reactivity of control rods in graphite moderated reactors

    International Nuclear Information System (INIS)

    Nakata, H.

    1978-01-01

    A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Gesh, Christopher J.

    2004-01-01

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

  3. Parametric study of moderator heat exchanger for Candu 6 advanced reactor

    International Nuclear Information System (INIS)

    Umar, Efrizon; Vecchiarelli, Jack

    2000-01-01

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

  4. The thorium fuel cycle in water-moderated reactor systems

    International Nuclear Information System (INIS)

    Critoph, E.

    1977-05-01

    Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An 'inventory' of uranium of between 1 and 2 Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium) is some two decades

  5. Organic materials for fusion-reactor applications

    International Nuclear Information System (INIS)

    Hurley, G.F.; Coltman, R.R. Jr.

    1983-09-01

    Organic materials requirements for fusion-reactor magnets are described with reference to the temperature, radiation, and electrical and mechanical stress environment expected in these magnets. A review is presented of the response to gamma-ray and neutron irradiation at low temperatures of candidate organic materials; i.e. laminates, thin films, and potting compounds. Lifetime-limiting features of this response as well as needed testing under magnet operating conditions not yet adequately investigated are identified and recomendations for future work are made

  6. Minor actinides incineration by loading moderated targets in fast reactor

    International Nuclear Information System (INIS)

    Wu Hongchun; Sato, Daisuke; Takeda, Toshikazu

    2000-01-01

    The effect of hydrogen concentration and loaded mass of minor actinides (MAs) in the target on the core performance and MAs transmutation rate was analyzed in this paper. An optimum core was proposed which has 96 MAs target assemblies of which MAs fuel pins per assembly is 38 with the composition ratio U/MA/Zr/H of 1/4/10/50. This optimized core offers good core performance and can transmute MAs very effectively, the transmutation rate was about 67% (939 kg) and the incinerate (transmute by fission) rate was about 35% (489 kg) through 3 years of reactor operation. It is about 2-3 times larger than current transmutation method that MAs are loaded homogeneously in the PWR and fast reactor core. (author)

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

    International Nuclear Information System (INIS)

    Aydogdu, K.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  11. Noise analysis method for monitoring the moderator temperature coefficient of pressurized water reactors: Neural network calibration

    International Nuclear Information System (INIS)

    Thomas, J.R. Jr.; Adams, J.T.

    1994-01-01

    A neural network was trained with data for the frequency response function between in-core neutron noise and core-exit thermocouple noise in a pressurized water reactor, with the moderator temperature coefficient (MTC) as target. The trained network was subsequently used to predict the MTC at other points in the same fuel cycle. Results support use of the method for operating pressurized water reactors provided noise data can be accumulated for several fuel cycles to provide a training base

  12. Organic food consumption in China: the moderating role of inertia

    Directory of Open Access Journals (Sweden)

    Yen Tsai-Fa

    2018-01-01

    Full Text Available Despite the progressive development of the organic food sector across Taiwan Strait, little is known about how consumers’ self congruity will influence organic food decision through various degrees of attitude and whether or not consumers with various degrees of inertia will vary in their intention to buy organic foods. The current study aims to examine the effect of consumption self congruity on behavioral intention related to organic food consumption under the mediating role of attitude as well as the moderating role of inertia. Research data were collected from organic food consumers across Taiwan Strait via a questionnaire survey, eventually obtaining 500 valid questionnaires for analysis. This study tested the overall model fit and hypotheses through structural equation modeling method (SEM. The results show that consumer attitude significantly mediates the effects of self congruity on organic food purchase intention. Moreover, the moderating effect of inertia is statistical significance, indicating that the relationship between attitude and purchase intention becomes weaker in the condition of consumers with higher degree of inertia. Several implications and suggestions are also discussed for organic food providers and marketers.

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

    International Nuclear Information System (INIS)

    Asaoka, Takumi; Watanabe, N.

    1976-01-01

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

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

    International Nuclear Information System (INIS)

    Carvalho Goncalves Junior, J. de.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Downar, T.J.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Bosevski, T.; Spiric, V.

    1964-01-01

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

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

    International Nuclear Information System (INIS)

    Less, T.J.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

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

    Directory of Open Access Journals (Sweden)

    Vorobiev Alexander V.

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi

    1982-12-01

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

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

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  7. Minor Actinides Burnup Enhancement in the European Sodium Fast Reactor through Moderator Material Addition

    International Nuclear Information System (INIS)

    Ramos, R.L.; Buiron, L.

    2013-01-01

    Conclusions: • ZrH 2 was the best moderator material, followed by MgO and MgAl 2 O 4 ; • When the number of moderator pins is increased: – the percentage of minor actinides consumed increases; – the total mass consumed of minor actinides decreases; – the decay heat generated decreases; – the neutron flux in the reactor varies very little. Perspectives: • For future studies it would be possible to evaluate the use of other materials with resonances in the scattering cross section in the fast range that would improve the results obtained with Mg. • It would be necessary to consider how to add moderator material without changing the initial mass of minor actinides. E.g., adding the moderator at the periphery of the minor actinide elements

  8. The rate of diffusion into advanced gas cooled reactor moderator bricks: an equivalent cylinder model

    International Nuclear Information System (INIS)

    Kyte, W.S.

    1980-01-01

    The graphite moderator bricks which make up the moderator of an advanced gas-cooled nuclear reactor (AGR) are of many different and complex shapes. Many physico-chemical processes that occur within these porous bricks include a diffusional step and thus to model these processes it is necessary to solve the diffusion equation (with chemical reaction) in a porous medium of complex shape. A finite element technique is applied to calculating the rate at which nitrogen diffuses into and out of the porous moderator graphite during operation of a shutdown procedure for an AGR. However, the finite element method suffers from several disadvantages that undermine its general usefulness for calculating rates of diffusion in AGR moderator cores. A model which overcomes some of these disadvantages is presented (the equivalent cylinder model) and it is shown that this gives good results for a variety of different boundary and initial conditions

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  10. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

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

    International Nuclear Information System (INIS)

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

    2000-06-01

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

  12. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  13. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  14. Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2003-01-01

    1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

  15. A moderation layer to improve the safety behavior of sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Weiß, F.P., E-mail: b.merk@fzd.de [Forschungszentrum Dresden-Rossendorf, Institut für Sicherheitsforschung, Dresden (germany)

    2011-07-01

    The nature of the sodium void effect in an infinite lattice is discussed and for a reduction of the effect the insertion of moderating material is proposed. The effect of three different moderating layers on the sodium void defect and the feedback effects is investigated. Especially the uranium zirconium hydride UzrH layer causes a strong reduction of the sodium void effect. Additionally, this layer improves the fuel temperature effect and the coolant effect of the system significantly. All changes caused by the insertion of the UZrH layer lead to a significant increase in stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. (author)

  16. A moderation layer to improve the safety behavior of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Merk, B.; Weiß, F.P.

    2011-01-01

    The nature of the sodium void effect in an infinite lattice is discussed and for a reduction of the effect the insertion of moderating material is proposed. The effect of three different moderating layers on the sodium void defect and the feedback effects is investigated. Especially the uranium zirconium hydride UzrH layer causes a strong reduction of the sodium void effect. Additionally, this layer improves the fuel temperature effect and the coolant effect of the system significantly. All changes caused by the insertion of the UZrH layer lead to a significant increase in stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. (author)

  17. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    International Nuclear Information System (INIS)

    Kansal, Anuj Kumar; Joshi, Jyeshtharaj B.; Maheshwari, Naresh Kumar; Vijayan, Pallippattu Krishnan

    2015-01-01

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated

  18. CFD analysis of moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kansal, Anuj Kumar, E-mail: akansal@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Maheshwari, Naresh Kumar, E-mail: nmahesh@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India); Vijayan, Pallippattu Krishnan, E-mail: vijayanp@barc.gov.in [Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)

    2015-06-15

    Highlights: • 3D CFD of vertical calandria vessel. • Spatial distribution of volumetric heat generation. • Effect of Archimedes number. • Non-dimensional analysis. - Abstract: Three dimensional computational fluid dynamics (CFD) analysis has been performed for the moderator flow and temperature fields inside a vertical calandria vessel of nuclear reactor under normal operating condition using OpenFOAM CFD code. OpenFOAM is validated by comparing the predicted results with the experimental data available in literature. CFD model includes the calandria vessel, calandria tubes, inlet header and outlet header. Analysis has been performed for the cases of uniform and spatial distribution of volumetric heat generation. Studies show that the maximum temperature in moderator is lower in the case of spatial distribution of heat generation as compared to that in the uniform heat generation in calandria. In addition, the effect of Archimedes number on maximum and average moderator temperature was investigated.

  19. Development of a silicon calorimeter for dosimetry applications in a water-moderated reactor

    International Nuclear Information System (INIS)

    DePriest, Kendall Russell; King, Donald Bryan; Naranjo, Gerald E.; Luker, Spencer Michael; Keltner, Ned R.; Suo-Anttila, Ahti Jorma; Griffin, Patrick Joseph

    2005-01-01

    High fidelity active dosimetry in the mixed neutron/gamma field of a research reactor is a very complex issue. For passive dosimetry applications, the use of activation foils addresses the neutron environment while the use of low neutron response CaF 2 :Mn thermoluminescent dosimeters (TLDs) addresses the gamma environment. While radiation-hardened diamond photoconducting detectors (PCD) have been developed that provide a very precise fast response (picosecond) dosimeter and can provide a time-dependent profile for the radiation environment, the mixed field response of the PCD is still uncertain and this interferes with the calibration of the PCD response. In order to address the research reactor experimenter's need for a dosimeter that reports silicon dose and dose rate at a test location during a pulsed reactor operation, a silicon calorimeter has been developed. This dosimeter can be used by itself to provide a dose in rad(Si) up to a point in a reactor pulsed operation, or, in conjunction with the diamond PCD, to provide a dose rate. This paper reports on the development, testing, and validation of this silicon calorimeter for applications in water-moderated research reactors.

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  1. Organization and processes of the BN-600 reactor mounting

    International Nuclear Information System (INIS)

    Dubrovin, E.Z.; Karpenko, V.N.; Takhtaulov, V.M.

    1982-01-01

    Structural peculiarities of the BN-600 reactor plant are considered. Experience of metal structure mounting inside the reactor vessel has been analysed. Recommendations on the improvements on the organization of the thermal mechanical equipment mounting are given. It is concluded that the consideration of these recommendations will permit to reduce expenditures of labour by 10-40% for the mounting

  2. Verification of codes used for the nuclear safety assessment of the small space heterogeneous reactors with zirconium hydride moderator

    International Nuclear Information System (INIS)

    Glushkov, E.S.; Gomin, E.A.; Kompaniets, G.V.

    1994-01-01

    Computer codes used for assessment of nuclear safety for space NPP are compared taking as an example small-sized heterogeneous reactor with zirconium hydride moderator of the Topaz-2 facility. The code verifications are made for five different variants

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-12-01

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

  4. Nitrogen Removal by Anammox Biofilm Column Reactor at Moderately Low Temperature

    Directory of Open Access Journals (Sweden)

    Tuty Emilia Agustina

    2017-10-01

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

  5. Method of collecting helium cover gas for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyamoto, Keiji; Ueda, Hiroshi.

    1981-01-01

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

  6. Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

    Directory of Open Access Journals (Sweden)

    Toshio Wakabayashi

    2013-01-01

    Full Text Available An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

  7. Improvements in gas supply systems for heavy-water moderated reactors

    International Nuclear Information System (INIS)

    Aubert, G.; Hassig, J.M.; Laurent, N.; Thomas, B.

    1964-01-01

    In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [fr

  8. Multidimensional space-time kinetics of a heavy water moderated nuclear reactor

    International Nuclear Information System (INIS)

    Winn, W.G.; Baumann, N.P.; Jewell, C.E.

    1980-01-01

    Diffusion theory analysis of a series of multidimensional space-time experiments is appraised in terms of the final experiment of the series. In particular, TRIMHX diffusion calculations were examined for an experiment involving free-fall insertion of a 235 U-bearing rod into a heavy water moderated reactor with a large reflector. The experimental transient flux-tilts were accurately reproduced after cross section adjustments forced agreement between static diffusion calculations and static reactor measurements. The time-dependent features were particularly well modeled, and the bulk of the small discrepancies in space-dependent features should be removable by more refined cross-section adjustments. This experiment concludes a series of space-time experiments that span a wide range of delayed neutron holdback effects. TRIMHX calculations of these experiments demonstrate the accuracy of the modeling employed in the code

  9. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Directory of Open Access Journals (Sweden)

    Seung Hyun Nam

    2015-10-01

    Full Text Available Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER, for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MWth and an electricity generation mode of 100 kWth, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  10. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW{sub th} and an electricity generation mode of 100 kW{sub th}, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and

  11. Innovative concept for an ultra-small nuclear thermal rocket utilizing a new moderated reactor

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Venneri, Paolo; Kim, Yong Hee; Lee, Jeong Ik; Chang, Soon Heung; Jeong, Yong Hoon

    2015-01-01

    Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for near-term human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of 100 MW th and an electricity generation mode of 100 kW th , equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics

  12. Fine distributed moderating material to the enhance feedback effects in LBE cooled rast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    In this work it is demonstrated, that the concept of enhanced feedback coefficients is transferable to LBE cooled fast reactors. The demonstration is based on the fuel assembly design of the CDT project. The effect of the moderating material on the neutron spectrum, on the k{sub inf}, and on the fuel temperature feedback and the coolant feedback is shown, discussed and compared to SFRs. The calculations are performed with the 2D lattice transport code HELIOS and based on the fully detailed fuel assembly geometry representation. (orig.)

  13. Contribution to the use of a solid moderator gas reactor, for naval propulsion

    International Nuclear Information System (INIS)

    Pheline, J.; Gautier, A.

    1960-01-01

    In this contribution, the authors discuss works performed in France for the development of nuclear propulsion in merchant ships, notably for an oil tanker of 50.000 tons with 17 knot speed, i.e. a 20.000 Hp engine with an energy produced by a 60 MW gas reactor with a solid moderator and comprising 400 channels loaded with uranium oxide enriched ay 2.8 per cent and sheathed with a refractory alloy. The authors discuss the possible materials for the moderator, the heat transfer medium, the sheath, the fuel and the structures, and report technological studies (mechanical tests, irradiation tests) performed to investigate material properties and their behaviour in operation conditions. They report tests performed to investigate core structure characteristics with respect to neutrons. They finally briefly present a prototype

  14. Three dimensional numerical simulation of a full scale CANDU reactor moderator to study temperature fluctuations

    International Nuclear Information System (INIS)

    Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc

    2014-01-01

    Highlights: • 3D model of a Candu reactor is modeled to investigate flow distribution. • The results show the temperature distribution is not symmetrical. • Temperature contours show the hot regions at the top left-hand side of the tank. • Interactions of momentum flows and buoyancy flows create circulation zones. • The results indicate that the moderator tank operates in the buoyancy driven mode. -- Abstract: Three dimensional numerical simulations are conducted on a full scale CANDU Moderator and transient variations of the temperature and velocity distributions inside the tank are determined. The results show that the flow and temperature distributions inside the moderator tank are three dimensional and no symmetry plane can be identified. Competition between the upward moving buoyancy driven flows and the downward moving momentum driven flows in the center region of the tank, results in the formation of circulation zones. The moderator tank operates in the buoyancy driven mode and any small disturbances in the flow or temperature makes the system unstable and asymmetric. Different types of temperature fluctuations are noted inside the tank: (i) large amplitude are at the boundaries between the hot and cold; (ii) low amplitude are in the core of the tank; (iii) high frequency fluctuations are in the regions with high velocities and (iv) low frequency fluctuations are in the regions with lower velocities

  15. Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Ibe, Eishi; Nakata, Kiyatomo; Fuse, Motomasa; Ohsumi, Katsumi; Takashima, Yoshie

    1995-01-01

    Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the crack growth rate, even if microcracks are present on the structural materials. Several disadvantage caused by HWC are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated base on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of Hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants -- e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom -- are planned that would collect data to support the theoretical considerations

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

    International Nuclear Information System (INIS)

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

    1977-11-01

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

  17. Physical properties of organic nuclear reactor coolants

    Energy Technology Data Exchange (ETDEWEB)

    Elberg, S.; Friz, G.

    1963-03-15

    Diphenyl and terphenyls with different high-boiler content were studied up to temperatures of 450 deg C. Data from high boiler reactors show viscosity (strong influence), thermal conductivity (medium influence), density and specific heat (small influence). The vapor pressure is rn the most affected property (important influence of low boilers). Also viscosity shows an effect. Some data for pure highboilers are also presented. New results were obtained with direct measurements of the latent heat ot vaporization. (P.C.H.)

  18. Study on Doppler coefficient for metallic fuel fast reactor added hydrogeneous moderator

    Energy Technology Data Exchange (ETDEWEB)

    Hirakawa, Naohiro; Iwasaki, Tomohiko; Tsujimoto, Kazuhumi [Tohoku Univ., Sendai (Japan). Faculty of Engineering; Osugi, Toshitaka; Okajima, Shigeaki; Andoh, Masaki; Nemoto, Tatsuo; Mukaiyama, Takehiko

    1998-01-01

    A series of mock-up experiments for moderator added metallic fast reactor core was carried out at FCA to obtain the experimental verification for improvement of reactivity coefficients. Softened neutron spectrum increases Doppler effect by a factor of 2, and flatter adjoint neutron spectrum decreases Na void effect by a factor of 0.6 when hydrogen to heavy metal atomic number ratio is increased from 0.02 to 0.13. The experimental results are analyzed with SLALOM and CITATION-FBR, which is the standard design code system for a fast reactor at JAERI, and SRAC95 and CITATION-FBR. The present code system gives generally good agreement with the experimental results, especially by the use of the latter, the dependence of the Doppler effect to the hydrogen to fuel element atomic number density ratio is disappeared. Therefore, it looks possible to use the present code system for the conceptual design of a fast reactor system with hydrogeneous materials. (author)

  19. Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

    2005-01-01

    R and D project to investigate thermal-hydraulic performance of tight-lattice fuel bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in collaboration with utilities, reactor vendors and universities from 2002. The RMWR realizes a high conversion ratio larger than 0.1 for sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The reactor core comprises tight-lattice fuel assemblies with gap clearance of around 1.0 mm to reduce the water volume ratio to achieve the high conversion ratio. A problem of utmost importance from a thermal-hydraulic point of view is the coolability of the tight-lattice assembly with such a small gap width. JAERI has been carrying out experimental study to investigate the system parameter effects on the thermal-hydraulic performance and to confirm the feasibility of the core. In the present study, the subchannel analysis code NASCA was applied to 37-rod tight-lattice bundle experiments. The NASCA can give good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy decreases for the gap width of 1.0 mm. To improve the prediction accuracy, the code will be modified to take the effect of film thickness distribution around fuel rods on boiling transition. (author)

  20. The organization of research reactor safety in the UKAEA

    International Nuclear Information System (INIS)

    Redpath, W.

    1983-01-01

    The present state of organization and development of research reactor safety in the UKAEA are outlined by addressing the fundamental safety principles which have been adopted in keeping with national health and safety requirement. The organisation, assessment and monitoring of research reactor safety on complex multi-discipline and multi-activity nuclear research and development site are discussed. Methods of safety assessment, such as probabilistic risk assessment and risk acceptance criteria, which have been developed and applied in practice are explained, and some indication of the directions in which some of the current developments in the safety of UKAEA research reactors is also included. (A.J.)

  1. Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zou, C.Y.; Cai, X.Z.; Jiang, D.Z.; Yu, C.G.; Li, X.X.; Ma, Y.W.; Han, J.L. [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Chen, J.G., E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

    2015-01-15

    Highlights: • The temperature feedback coefficient with different moderation ratios for TMSR in thermal neutron region is optimized. • The breeding ratio and doubling time of a thermal TMSR with three different reprocessing schemes are analyzed. • The smaller hexagon size and larger salt fraction with more negative feedback coefficient can better satisfy the safety demands. • A shorter reprocessing time can achieve a better breeding ratio in a thermal TMSR. • The graphite moderator lifespan is compared with other MSRs and discussed. - Abstract: Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed.

  2. 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

  3. Sustainability management for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  4. Sustainability management for operating organizations of research reactors

    International Nuclear Information System (INIS)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de

    2017-01-01

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

  5. Effects of cryogenic reactor irradiation on organic insulators

    International Nuclear Information System (INIS)

    Kato, Teruo

    1995-01-01

    Insulators for the superconducting magnets of fusion reactor are classified as electrical and thermal insulators for which tough organic materials will be used. When the magnet is exposed by fast neutrons and gamma-rays from plasma in a fusion reactor, the fusion reactor systems will cause fatal damage by the degradation of insulators. Therefore, it is necessary to select materials resistant irradiation damage for use as insulators. Electrical and mechanical tests were carried out at 4.2 K without warmup after the reactor irradiation at 5 K. The effects of reactor irradiation at the dose of 10 7 Gy on epoxy resins (bisphenol-A), G-10 CR, VL-E 200 and G-11 CR caused large decreases in mechanical strength. Polyetheretherketone (PEEK), polyimide and phenol novolac resins, which were used to laminate reinforced plastics with glass-cloth against irradiation, showed good resistance. Effects of cryogenic reactor irradiation on several organic materials and epoxy laminate-reinforced plastics with glass-cloth and Kevlar-cloth were also discussed. (author)

  6. Research and development on reduced-moderation light water reactor with passive safety features (Contract research)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Akie, Hiroshi; Kugo, Teruhiko; Yonomoto, Taisuke; Kureta, Masatoshi; Ishikawa, Nobuyuki; Nagaya, Yasunobu; Araya, Fumimasa; Okajima, Shigeaki; Okumura, Keisuke; Suzuki, Motoe; Mineo, Hideaki; Nakatsuka, Toru

    2004-06-01

    The present report contains the achievement of 'Research and Development on Reduced-moderation Light Water Reactor with Passive Safety Features', which was performed by Japan Atomic Energy Research Institute (JAERI), Hitachi Ltd., Japan Atomic Power Company and Tokyo Institute of Technology in FY2000-2002 as the innovative and viable nuclear energy technology (IVNET) development project operated by the Institute of Applied Energy (IAE). In the present project, the reduced-moderation water reactor (RMWR) has been developed to ensure sustainable energy supply and to solve the recent problems of nuclear power and nuclear fuel cycle, such as economical competitiveness, effective use of plutonium and reduction of spent fuel storage. The RMWR can attain the favorable characteristics such as high burnup, long operation cycle, multiple recycling of plutonium (Pu) and effective utilization of uranium resources based on accumulated LWR technologies. Our development target is 'Reduced-moderation Light Water Reactor with Passive Safety Features' with innovative technologies to achieve above mentioned requirement. Electric power is selected as 300 MWe considering anticipated size required for future deployment. The reactor core consists of MOX fuel assemblies with tight lattice arrangement to increase the conversion ratio. Design targets of the core specification are conversion ratio more than unity, negative void reactivity feedback coefficient to assure safety, discharged burnup more than 60 GWd/t and operation cycle more than 2 years. As for the reactor system, a small size natural circulation BWR with passive safety systems is adopted to increase safety and reduce construction cost. The results obtained are as follows: As regards core design study, core design was performed to meet the goal. Sequence of startup operation was constructed for the RMWR. As the plant design, plant system was designed to achieve enhanced economy using passive safety system effectively. In

  7. Pressure drop characteristics in tight-lattice bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime

    2004-01-01

    The reduced-moderation water reactor (RMWR) consists of several distinctive structures; a triangular tight-lattice configuration and a double-flat core. In order to design the RMWR core from the point of view of thermal-hydraulics, an evaluation method on pressure drop characteristics in the rod bundles at the tight-lattice configuration is required. In this study, calculated results by the Martinelli-Nelson's and Hancox's correlations were compared with experimental results in 4 x 5 rod bundles and seven-rod bundles. Consequently, the friction loss in two-phase flows becomes smaller at the tight-lattice configuration with the hydraulic diameter less than about 3 mm. This reason is due to the difference of the configuration between the multi-rod bundle and the circular tube and due to the effect of the small hydraulic diameter on the two-phase multiplier. (author)

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

    International Nuclear Information System (INIS)

    Joemon, V.; Sinha, R.K.

    1993-01-01

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

  9. Optimization of seed-blanket type fuel assembly for reduced-moderation water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shelley, Afroza; Shimada, Shoichiro; Kugo, Teruhiko; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi

    2003-10-01

    Parametric studies have been performed for a PWR-type reduced-moderation water reactor (RMWR) with the seed-blanket type fuel assembles to achieve a high conversion ratio, negative void reactivity coefficient and a high burnup by using MOX fuel. From the viewpoint of reactor safety analysis, the fuel temperature coefficients were also studied. From the result of the burnup calculation, it has been seen that ratio of 40-50% of outer blanket in a seed-blanket assembly gives higher conversion ratio compared to the other combination of seed-blanket assembly. And the recommended number of (seed+blanket) layers is 20, in which the number of seed (S) layers is 15 (S15) and blanket (B) layers is 5 (B5). It was found that the conversion ratio of seed-blanket assembly decreases, when they are arranged looks like a flower shape (Hanagara). By the optimization of different parameters, S15B5 fuel assembly with the height of seed of 1000 mmx2, internal blanket of 150 mm and axial blanket of 400 mmx2 is recommended for a reactor of high conversion ratio. In this assembly, the gap of seed fuel rod is 1.0 mm and blanket fuel rod is 0.4 mm. In S15B5 assembly, the conversion ratio is 1.0 and the burnup is 38.18 GWd/t in (seed+internal blanket+outer blanket) region. However, the burnup is 57.45 GWd/t in (seed+internal blanket) region. The cycle length of the core is 16.46 effective full power in month (EFPM) by six batches and the enrichment of fissile Pu is 14.64 wt.%. The void coefficient is +21.82 pcm/%void, however, it is expected that the void coefficient will be negative if the radial neutron leakage is taken into account in the calculation. It is also possible to use S15B5 fuel assembly as a high burnup reactor 45 GWd/t in (seed+internal blanket+outer blanket) region, however, it is necessary to decrease the height of seed to 500 mmx2 to improve the void coefficient. In this reactor, the conversion ratio is 0.97 and void coefficient is +20.81 pcm/%void. The fuel temperature

  10. Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2006-01-01

    As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

  11. Reactor physics measurements with 19-element ThOsub(2)-sup(235)UOsub(2) cluster fuel in heavy water moderator

    International Nuclear Information System (INIS)

    French, P.M.

    1985-02-01

    Low power lattice physics measurements have been performed with a single rod of 19-element thorium oxide fuel enriched with 1.45 wt. percent sub(235)UOsub(2) (93 percent enriched) in a simulated heavy water moderated and cooled power reactor core. The experiments were designed to provide data relevant to a power reactor irradiation and to obtain some basic information on the physics of uranium-thorium fuel material. Some theoretical flux calculations are summarized and show reasonable agreement with experiment

  12. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  13. Energetics of semi-catalyzed-deuterium, light-water-moderated, fusion-fission toroidal reactors

    International Nuclear Information System (INIS)

    Jassby, D.L.; Towner, H.H.; Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.

    1978-07-01

    The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U 3 Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10 14 cm -3 s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor

  14. 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

  15. Startup of a high-temperature reactor cooled and moderated by supercritical-pressure light water

    International Nuclear Information System (INIS)

    Yi, Tin Tin; Ishiwatari, Yuki; Koshizuka, Seiichi; Oka, Yoshiaki

    2003-01-01

    The startup schemes of high-temperature reactors cooled and moderated by supercritical pressure light water (SCLWR-H) with square lattice and descending flow type water rods are studied by thermal-hydraulic analysis. In this study, two kinds of startup systems are investigated. In the constant pressure startup system, the reactor starts at a supercritical pressure. A flash tank and pressure reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at a subcritical pressure. A steam-water separator and a drain tank are required for two-phase flow at startup. The separator is designed by referring to the water separator used in supercritical fossil-fired power plants. The maximum cladding surface temperature during the power-raising phase of startup is restricted not to exceed the rated value of 620degC. The minimum feedwater flow rate is 25% for constant pressure startup and 35% for sliding pressure startup system. It is found that both constant pressure startup system and sliding pressure startup system are feasible in SCLWR-H from the thermal hydraulic point of view. The core outlet temperature as high as 500degC can be achieved in the present design of SCLWR-H. Since the feedwater flow rate of SCLWR-H (1190 kg/s) is lower than that of the previous SCR designs the weight of the component required for startup is reduced. The sliding pressure startup system is better than constant pressure startup system in order to reduce the required component weight (and hence material expenditure) and to simplify the startup plant system. (author)

  16. Organization Tenure as a Moderator of the Job Satisfaction-Job Performance Relationship.

    Science.gov (United States)

    Norris, Dwight R.; Niebuhr, Robert E.

    1984-01-01

    Examined organization tenure as a moderator of the job satisfaction-job performance relationship with technical employees (N=116) in a medium-sized industrial firm. Results provided support for a moderating influence of organization tenure on the relationship between job performance and overall satisfaction, satisfaction with work, and…

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. Heavy water moderated gas-cooled reactors; Filiere eau lourde - gaz

    Energy Technology Data Exchange (ETDEWEB)

    Bailly du Bois, B; Bernard, J L; Naudet, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

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

  20. Production of organic acids in an immobilized cell reactor using ...

    African Journals Online (AJOL)

    Immobilized cell reactor (ICR) was developed as a novel bioreactor to convert hydrolyzed sugars to organic acids. Sugar fermentation by Propionibacterium acid-propionici entraped by calcium alginate was carried out in continuous mode to produce propionic and acetic acids. In continuous fermentation, more than 90 ...

  1. Degradation Mechanisms of Colloidal Organic Matter in Biofilm Reactors

    DEFF Research Database (Denmark)

    Larsen, Tove; Harremoës, Poul

    1994-01-01

    -diffusible organic matter in a biofilm reactor. DH depends on the combined volumetric and surface hydraulic loading rate, Q2/(AV). In full-scale wastewater treatment plants, the degradation mechanism presented in this paper can explain important differences between the performance of trickling filters and RBC...

  2. Design of a graphite-moderated {sup 241}Am-Li neutron field to simulate reactor spectra

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, N., E-mail: tsujimura.norio@jaea.go.j [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Yoshida, T. [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan)

    2010-12-15

    A neutron calibration field using {sup 241}Am-Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the {sup 241}Am-Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.

  3. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  5. Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

    International Nuclear Information System (INIS)

    High, H.M.

    1983-05-01

    The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

  6. Scaling of silent electrical discharge reactors for hazardous organics destruction

    International Nuclear Information System (INIS)

    Coogan, J.J.; Rosocha, L.A.; Brower, M.J.; Kang, M.; Schmidt, C.A.

    1993-01-01

    Silent electrical discharges are used to produce highly reactive free radicals that destroy hazardous compounds entrained in gaseous effluents at ambient gas temperatures and pressures. We have carried out destruction experiments at Los Alamos on a range of volatile organic compounds (VOCs), including trichloroethylene (TCE), carbon tetrachloride, perchloroethylene (PCE), and chlorofluorocarbons (CFCs). We have measured a ''nine-factor'', the amount of energy required to reduce the VOC concentration by a factor of ten. For practical reactor power densities, the ''nine-factor'' can be used to predict the destruction an removal efficiency (DRE) in terms of gas flow rate and the number of reactor modules. This report proposes a modular, stackable architecture for scaling up the reactor throughput

  7. Safety analysis of high temperature reactor cooled and moderated by supercritical light water

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

    2003-01-01

    This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint

  8. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    International Nuclear Information System (INIS)

    Fukaya, Y.; Okubo, T.; Uchikawa, S.

    2008-01-01

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the 241 Pu content in the initial fuel, and the decay heat mainly depends on 238 Pu and 244 Cm. The contribution of 244 Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum. In addition, from

  9. Experiences in the emptying of waste silos containing solid nuclear waste from graphite- moderated reactors

    International Nuclear Information System (INIS)

    Wall, S.; Schwarz, T.

    2003-01-01

    Before reactor sites can be handed over for ultimate decommissioning, at some sites silos containing waste from operations need to be emptied. The form and physical condition of the waste demands sophisticated retrieval technologies taking into account the onsite situation in terms of infrastructure and silo geometry. Furthermore, in the case of graphite moderated reactors, this waste usually includes several tonnes of graphite waste requiring special HVAC and dust handling measures. RWE NUKEM Group has already performed several contracts dealing with such emptying tasks. Of particular interest for the upcoming decommissioning projects in France might be the activities at Vandellos, Spain and Trawsfynnyd, UK. Retrieval System for Vandellos NPP is discussed. Following an international competitive tender exercise, RWE NUKEM won the contract to provide a turn-key retrieval system. This involved the design, manufacture and installation of a system built around the modules of a 200 kg capacity version of the ARTISAN manipulator system. The ARTISAN 200 manipulator, with remote slave arm detach facility, was deployed on a telescopic mast inserted into the silos through the roof penetrations. The manipulator deployed a range of tools to gather the waste and load it into a transfer basket, deployed through an adjacent penetration. After commissioning, the system cleared the vaults in less than the scheduled period with no failures. At the Trawsfynnyd Magnox plants two types of intermediate level waste (ILW) accumulated on site; namely Miscellaneous Activated Components (MAC) and Fuel Element Debris (FED). MAC is predominantly components that have been activated by the reactor core and then discharged. FED mainly consists of fuel cladding produced when fuel elements were prepared for dispatch to the reprocessing facility. RWE NUKEM Ltd. was awarded a contract to design, supply, commission and operate equipment to retrieve, pack and immobilize the two waste streams. Major

  10. Investigation on spent fuel characteristics of reduced-moderation water reactor (RMWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fukaya, Y. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)], E-mail: fukaya.yuji@jaea.go.jp; Okubo, T.; Uchikawa, S. [Advanced Nuclear System Research and Development Directorate, Japan Atomic Energy Agency (JAEA), Oarai-machi, Ibaraki-ken 311-1393 (Japan)

    2008-07-15

    The spent fuel characteristics of the reduced-moderation water reactor (RMWR) have been investigated using the SWAT and ORIGEN codes. RMWR is an advanced LWR concept for plutonium recycling by using the MOX fuel. In the code calculation, the ORIGEN libraries such as one-group cross-section data prepared for RMWR were necessary. Since there were no open libraries for RMWR, they were produced in this study by using the SWAT code. New libraries based on the heterogeneous core modeling in the axial direction and with the variable actinide cross-section (VXSEC) option were produced and selected as the representative ORIGEN libraries for RMWR. In order to investigate the characteristics of the RMWR spent fuel, the decay heat, the radioactivity and the content of each nuclide were evaluated with ORIGEN using these libraries. In this study, the spent fuel characteristics of other types of reactors, such as PWR, BWR, high burn-up PWR, full-MOX-PWR, full-MOX-BWR and FBR, were also evaluated with ORIGEN. It has been found that about a half of the decay heat of the RMWR spent fuel comes from the actinides nuclides. It is the same with the radioactivity. The decay heat and the radioactivity of the RMWR spent fuel are lower than those of full-MOX-LWRs and FBR, and are the same level as those of the high burn-up PWR. The decay heat and the radioactivity from the fission products (FPs) in the spent fuel mainly depend on the burn-up and the burn-up time rather than the reactor type. Therefore, the decay heat and the radioactivity from FPs in the RMWR spent fuel are smaller, reflecting its relatively long burn-up time resulted from its core characteristics with the high conversion ratio. The radioactivity from the actinides in the spent fuel mainly depends on the {sup 241}Pu content in the initial fuel, and the decay heat mainly depends on {sup 238}Pu and {sup 244}Cm. The contribution of {sup 244}Cm is much smaller in RMWR than in MOX-LWRs because of the difference in the spectrum

  11. A Conceptual Supercritical Water Cooled Reactor Design Using a Cruciform Solid Moderator

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Bae, Kang Mok; Yoo, Jae Woon; Lee, Hyun Chul; Noh, Jae Man; Bae, Yoon Yong

    2006-01-15

    A Super Critical Water-Cooled Reactor(SCWR) concept proposed by Gen-IV has an advantage of a high thermal efficiency. However, there are some difficulties in neutronic core design for a SCWR due to lower moderator density resulting from the high operating temperature over the pseudo-critical temperature. In this report, the design concepts for the fuel assembly and the core for a SCWR were described as a feasibility study on the SCWR core design. HELIOS lattice code which will be used for group constants generation was verified for the application to the low coolant density condition of a SCWR. The TAF module for a thermal hydraulic feedback in MASTER was modified to consider high pressure and temperature of the supercritical coolant with single-phase fluid. A cruciform ZrH{sub 2} solid moderator was proposed for the SCWR fuel assembly design to compensate the lower coolant density. The axial zoning concept with three different enrichments for a fuel rod was used for the axial power shape control. Gadolinia burnable poison rods were used to reduce excess reactivity. Control rod system was grouped into 6 banks to control the excess reactivity of the core during normal operation. An orifice concept for each assembly was applied to control a coolant flow rate individually. As a result of the neutronic analysis for the equilibrium SCWR core, the maximum linear heat generation rete limit was satisfied and the maximum coolant temperature of the core outlet was {approx}590 .deg. C which is lower than 620 .deg. C of the maximum clad temperature limit.

  12. A flashing driven moderator cooling system for CANDU reactors: Experimental and computational results

    International Nuclear Information System (INIS)

    Khartabil, H.F.

    2000-01-01

    A flashing-driven passive moderator cooling system is being developed at AECL for CANDU reactors. Preliminary simulations and experiments showed that the concept was feasible at normal operating power. However, flow instabilities were observed at low powers under conditions of variable and constant calandria inlet temperatures. This finding contradicted code predictions that suggested the loop should be stable at all powers if the calandria inlet temperature was constant. This paper discusses a series of separate-effects tests that were used to identify the sources of low-power instabilities in the experiments, and it explores methods to avoid them. It concludes that low-power instabilities can be avoided, thereby eliminating the discrepancy between the experimental and code results. Two factors were found to be important for loop stability: (1) oscillations in the calandria outlet temperature, and (2) flashing superheat requirements, and the presence of nucleation sites. By addressing these factors, we could make the loop operate in a stable manner over the whole power range and we could obtain good agreement between the experimental and code results. (author)

  13. The emergency organization of the Federal Institute for Reactor Research

    International Nuclear Information System (INIS)

    Brunner, H.; Huerlimann, T.

    1977-01-01

    The organization and means of the emergency organization of the Federal Institute for Reactor Research (EIR), the eldest and largest nuclear installation in Switzerland, are described. It consists of a central command group and the following emergency teams: fire brigade, radiation protection, first aid, control, operating teams (reactors, hot laboratory etc.). The radiation protection team is formed by the Health Physics Division and is discussed in detail. A description of the alarm system and the first actions to be taken in case of an emergency is given. The importance of frequent and well-planned exercises and of radio communication between the teams and the command group is stressed and the emergency training programme of the EIR School for Radiation Protection, operated by the Health Physics Division, is presented. A fortunate lack of incidents at EIR is partly compensated for by experience gained from emergency team assistance operations during incidents outside the Institute. (author)

  14. The moderator's moderator

    International Nuclear Information System (INIS)

    Williamson, G.K.

    1990-01-01

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

  15. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Boldyrev, V.M.; Mikhaj, V.I.

    1985-01-01

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  16. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    International Nuclear Information System (INIS)

    Chang, Se Myong; Park, A. Y.; Kim, Hyoung Tae

    2011-01-01

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  17. Thermo-hydraulic test of the moderator cell of liquid hydrogen cold neutron source for the Budapest research reactor

    International Nuclear Information System (INIS)

    Grosz, Tamas; Rosta, Laszlo; Hargitai, Tibor; Mityukhlyaev, V.A.; Serebrov, A.P.; Zaharov, A.A.

    1999-01-01

    Thermo-hydraulic experiment was carried out in order to test performance of the direct cooled liquid hydrogen moderator cell to be installed at the research reactor of the Budapest Neutron Center. Two electric hearers up to 300 W each imitated the nuclear heat release in the liquid hydrogen as well as in construction material. The test moderator cell was also equipped with temperature gauges to measure the hydrogen temperature at different positions as well as the inlet and outlet temperature of cooling he gas. The hydrogen pressure in the connected buffer volume was also controlled. At 140 w expected total heat load the moderator cell was filled with liquid hydrogen within 4 hours. The heat load and hydrogen pressure characteristics of the moderator cell are also presented. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  20. Unique differences in applying safety analyses for a graphite moderated, channel reactor

    International Nuclear Information System (INIS)

    Moffitt, R.L.

    1993-06-01

    Unlike its predecessors, the N Reactor at the Hanford Site in Washington State was designed to produce electricity for civilian energy use as well as weapons-grade plutonium. This paper describes the major problems associated with applying safety analysis methodologies developed for commercial light water reactors (LWR) to a unique reactor like the N Reactor. The focus of the discussion is on non-applicable LWR safety standards and computer modeling/analytical variances of standards. The approaches used to resolve these problems to develop safety standards and limits for the N Reactor are described

  1. The feasibility study of using deuterated gadolinium nitrate for moderator-poisoned shutdown and excess reactivity control in CANDU reactors

    International Nuclear Information System (INIS)

    Li, J.; Everatt, A.

    2006-01-01

    Gadolinium nitrate is used in CANDU stations as moderator poison for reactor shutdowns and excess reactivity control. The use of the light-water hydrate introduces significant quantities of light water into the moderator system, which must be removed from the moderator by periodically upgrading the moderator (isotopic maintenance). The benefit of using a deuterated gadolinium nitrate would be a higher moderator isotopic and/or a lesser isotopic maintenance requirement. This study evaluated the economics of using deuterated gadolinium nitrate, as opposed to the light-water hydrate, for moderator-poisoned shutdowns and excess reactivity control in CANDU-6 reactors. Normal gadolinium nitrate (i.e., the light-water hydrate) is available from suppliers at ∼125 $/kg. Supplier quotes for deuterated gadolinium nitrate ranged from 1900 to 4000 $/kg. To examine the possibility of producing deuterated gadolinium nitrate in-house at a lower cost than commercially available, a three-stage dissolution/evaporation manufacturing process was conceived and costed. Depending on the assumed demand for the product (i.e., the number of reactors adopting the use of the product) and the capital recovery period, the estimated unit cost for the dissolution/evaporation process ranged from 730 to 2500 $/kg. The determination of economic benefit from using deuterated gadolinium nitrate in existing CANDU stations was based on the cost savings resulting from a higher fuel burn-up (i.e., the higher moderator isotopic would give a higher fuel burn-up). The net benefit of using deuterated gadolinium nitrate for most CANDU stations was determined to be marginal (i.e., <20 k$/a). Only for those CANDU stations where the moderator isotopic was relatively low (e.g., 99.85 wt%) was there a potential significant benefit (20-100 k$/a). However, if the reason for the low moderator isotopic is a relatively high moderator light-water ingress rate from sources other than the use of the light-water hydrate

  2. On the use of a moderation layer to improve the safety behavior in sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno, E-mail: b.merk@fzd.de [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany); Fridman, Emil; Weiss, Frank-Peter [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany)

    2011-05-15

    Research highlights: > Using a moderation layer can reduce the sodium void effect in a SFR. > Inserting the moderation layer improves the Doppler effect significantly. > The uniform layer distribution avoids effects on power and burnup distribution. > Hydride containing material like uranium-zirconium hydride is most efficient. - Abstract: This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B{sub 4}C or uranium-zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.

  3. Study on neutronics performance of flower shape advanced supercritical water cooled fast reactor with different solid moderators

    International Nuclear Information System (INIS)

    Yu Tao; Li Zhifeng; Xie Jinsen; Peng Honghua

    2015-01-01

    The supercritical water cooled fast reactors worked at such harsh condition with high temperature and high pressure, huge hydrogen balance pressure and thermal shock can result in a great loss of hydrogen. The released hydrogen would be out of control under accident situations. K_e_f_f, conversion ratio, moderator temperature effect, Doppler effect and void effect of different material such as ZrH_1_._7, Bp, BeO, C and SiC are discussed. BeO and SiC hold better integrated performance among these materials. Besides, moderators have less effect on the Doppler effect of fuel. (authors)

  4. Study of organic waste for production of hydrogen in reactor

    International Nuclear Information System (INIS)

    Guzmán Chinea, Jesús Manuel; Guzmán Marrero, Elizabeth; Pérez Ponce, Alejandro

    2015-01-01

    Biological processes have long been used for the treatment of organic waste makes, especially our study is based on the anaerobic process in reactors, using residual organic industry. Without excluding other non-industrial we have studied. Fundamental objectives treating organic waste is to reduce the pollutant load to the environment, another aim is to recover the waste recovering the energy contained in it. In this context, the biological hydrogen production from organic waste is an interesting alternative because it has low operating costs and raw material is being used as a residue in any way should be treated before final disposal. Hydrogen can be produced sustainable by anaerobic bacteria that grow in the dark with rich carbohydrate substrates giving as final products H 2 , CO 2 and volatile fatty acids. The whey byproduct from cheese production, has great potential to be used for the generation of hydrogen as it has a high carbohydrate content and a high organic load. The main advantages of using anaerobic processes in biological treatment of organic waste, are the low operating costs, low power consumption, the ability to degrade high organic loads, resistance biomass to stay long in the absence of substrate, without lose their metabolic activity, and low nutritional requirements and increase the performance of 0.9 mol H2 / mol lactose. (full text)Biological processes have long been used for the treatment of organic waste makes, especially our study is based on the anaerobic process in reactors, using residual organic industry. Without excluding other non-industrial we have studied. Fundamental objectives treating organic waste is to reduce the pollutant load to the environment, another aim is to recover the waste recovering the energy contained in it. In this context, the biological hydrogen production from organic waste is an interesting alternative because it has low operating costs and raw material is being used as a residue in any way should be treated

  5. 10-75-kWe-reactor-powered organic Rankine-cycle electric power systems (ORCEPS) study. Final technical report

    Energy Technology Data Exchange (ETDEWEB)

    1977-03-30

    This 10-75 kW(e) Reactor-ORCEPS study was concerned with the evaluation of several organic Rankine cycle energy conversion systems which utilized a /sup 235/U-ZrH reactor as a heat source. A liquid metal (NaK) loop employing a thermoelectric converter-powered EM pump was used to transfer the reactor energy to the organic working fluid. At moderate peak cycle temperatures (750/sup 0/F), power conversion unit cycle efficiencies of up to 25% and overall efficiencies of 20% can be obtained. The required operating life of seven years should be readily achievable. The CP-25 (toluene) working fluid cycle was found to provide the highest performance levels at the lowest system weights. Specific weights varies from 100 to 50 lb/kW(e) over the power level range 10 to 75 kW(e). (DLC)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  7. Effective neutron temperature measurements in well moderated reactor by the reactivity coefficient method

    International Nuclear Information System (INIS)

    Raisic, N.; Klinc, T.

    1968-11-01

    The ratio of the reactivity changes of a nuclear reactor produced by successive introduction of two different neutron absorbers in the reactor core, has been measured and information on effective neutron temperature at a particular point obtained. Boron was used as a l/v absorber and cadmium as an absorber sensiti ve to neutron temperature. Effective neutron temperature distribution has been deduced by moving absorbers across the reactor core and observing the corresponding reactivity changes. (author)

  8. Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

    International Nuclear Information System (INIS)

    Shunsuke Uchida; Eishi Ibe; Katsumi Ohsumi

    1994-01-01

    Application of hydrogen water chemistry to moderate corrosive circumstances is a promising approach to preserve structural integrities of major components and structures in the primary cooling system of BWRs. The benefits of HWC application are usually accompanied by several disadvantages. After evaluating merits and demerits of HWC application, it is concluded that optimal amounts of hydrogen injected into the feed water can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. (authors). 1 fig., 4 refs

  9. Analysis of passive moderator cooling system of Candu-6A reactor at emergency condition

    International Nuclear Information System (INIS)

    Umar, Efrizon; Subki, M. Hadid; Vecchiarelli, Jack

    2001-01-01

    Analysis of passive moderator cooling system subject to in-core LOCA with no emergency core cooling injection has been done. In this study, the new model of passive moderator system has been tested for emergency conditions and CATHENA code Mod-3.5b/Rev1 is used to calculate some parameters of this passive moderator cooling system. This result of simulation show that the proposed moderator cooling system have given satisfactory result, especially for the case with 0.7 m riser diameter and the number of heat exchanger tubes 8100. For PEWS tank containing 3000 m3 of light water initially at 30 0C and a 3641 m2 moderator heat exchanger, the average long-term heat removed rate balances the moderator heat load and the flow through the passive moderator loop remains stable for over 72 hours with no saturated boiling in the calandria and flow instabilities do not develop during long-term period

  10. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

  11. Aiming at super long term application of nuclear energy. Scope and subjects on the water cooled breeder reactor, the 'reduced moderation water reactor'

    International Nuclear Information System (INIS)

    Sato, Osamu; Tatematsu, Kenji; Tanaka, Yoji

    2001-01-01

    In order to make possible on nuclear energy application for super long term, development of sodium cooling type fast breeder reactor (FBR) has been carried out before today. However, as it was found that its commercialization was technically and economically difficult beyond expectation, a number of nations withdrew from its development. And, as Japan has continued its development, scope of its actual application is not found yet. Now, a research and development on a water cooling type breeder reactor, the reduced moderation water reactor (RMWR)' using LWR technology has now been progressed under a center of JAERI. This RMWR is a reactor intending a jumping upgrade of conversion ratio by densely arranging fuel bars to shift neutron spectrum to faster region. The RMWR has a potential realizable on full-dress plutonium application at earlier timing through its high conversion ratio, high combustion degree, plutonium multi-recycling, and so on. And, it has also feasibility to solve uranium resource problem by realization of conversion ratio with more than 1.0, to contribute to super long term application of nuclear energy. Here was investigated on an effect of reactor core on RMWR, especially of its conversion ratio and plutonium loading on introduction effect as well as on how RMWR could be contributed to reduction of uranium resource consumption, by drawing some scenario on development of power generation reactor and fuel cycle in Japan under scope of super long term with more than 100 years in future. And, trial calculation on power generation cost of the RMWR was carried out to investigate some subjects at a viewpoint of upgrading on economy. (G.K.)

  12. Development of the heavy-water organic-cooled reactor. Status report from the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Trilling, C A [Atomics International, Division of North American Aviation, Inc., Canoga Park, CA (United States)

    1967-01-01

    loops required. A second benefit of the pressure tube concept is that it is easily adaptable to the use of on-power refuelling, with the consequent potential of achieving high plant availability and optimizing the fuel management programme for maximum neutron economy. Many studies have been carried out in an attempt to optimize the selection of the coolant for a heavy-water-moderated reactor. The choice of an organic coolant for the US heavy-water reactor development programme offers several advantages. By limiting the heavy water to its function as moderator, its containment is required only at low temperatures and pressures, thus minimizing both inventory and losses of this expensive material. The compatibility of the organic coolant with uranium, plutonium, and thorium metals, and with their oxides and carbides provides for maximum flexibility in the selection of fuel material for this reactor. The HWOCR can therefore take full advantage of whichever fuel cycle in the long run demonstrates the most favourable economics. The low vapour pressure of the organic coolant and its compatibility with standard materials of construction provide for the design of low pressure primary coolant loops using carbon and low alloy steels while obtaining the thermodynamic efficiency of a superheat steam cycle. The low level of induced radioactivity normally present in the organic coolant permits normal contact maintenance of the primary coolant loops while the plant is in operation.

  13. Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

    Energy Technology Data Exchange (ETDEWEB)

    Dris, Zakaria bin, E-mail: zakariadris@gmail.com [College of Graduate Studies, Universiti Tenaga Nasional (UNITEN), Putrajaya Campus, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Centre for Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Putrajaya Campus, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Mohamed, Abdul Aziz bin; Hamid, Nasri A. [Centre for Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Putrajaya Campus, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Azman, Azraf; Ahmad, Megat Harun Al Rashid Megat; Jamro, Rafhayudi; Yazid, Hafizal [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried out using a neutron spectrometer.

  14. Chapter 10: Calculation of the temperature coefficient of reactivity of a graphite-moderated reactor

    International Nuclear Information System (INIS)

    Brown, G.; Richmond, R.; Stace, R.H.W.

    1963-01-01

    The temperature coefficients of reactivity of the BEPO, Windscale and Calder reactors are calculated, using the revised methods given by Lockey et al. (1956) and by Campbell and Symonds (1962). The results are compared with experimental values. (author)

  15. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

  16. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

    2005-01-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  17. State of the art of nuclear facilities with organic cooled reactors

    International Nuclear Information System (INIS)

    Brede, O.

    1984-01-01

    USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined. (author)

  18. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor

    International Nuclear Information System (INIS)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J.

    2004-01-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  19. Application of the dose limitation system to the control of carbon-14 releases from heavy-water-moderated reactors

    International Nuclear Information System (INIS)

    Beninson, D.; Gonzalez, A.J.

    1982-01-01

    Heavy-water-moderated reactors produce substantially more carbon-14 than light-water reactors. Applying the principles of the systems of dose limitation, the paper presents the rationale used for establishing the release limit for effluents containing this nuclide and for the decisions made regarding the effluent treatment in the third nuclear power station in Argentina. Production of carbon-14 in PHWR and the release routes are analysed in the light of the different effluent treatment possibilities. An optimization assessment is presented, taking into account effluent treatment and waste management costs, and the collective effective dose commitment due to the releases. The contribution of present carbon-14 releases to future individual doses is also analysed in the light of an upper bound for the contribution, representing a fraction of the individual dose limits. The paper presents the resulting requirements for the effluent treatment regarding carbon-14 and the corresponding regulatory aspects used in Argentina. (author)

  20. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    International Nuclear Information System (INIS)

    Lindley, Benjamin A.; Parks, Geoffrey T.; Franceschini, Fausto

    2013-01-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  1. Transmutation, Burn-Up and Fuel Fabrication Trade-Offs in Reduced-Moderation Water Reactor Thorium Fuel Cycles - 13502

    Energy Technology Data Exchange (ETDEWEB)

    Lindley, Benjamin A.; Parks, Geoffrey T. [University of Cambridge, Cambridge (United Kingdom); Franceschini, Fausto [Westinghouse Electric Company LLC, Cranberry Township, PA (United States)

    2013-07-01

    Multiple recycle of long-lived actinides has the potential to greatly reduce the required storage time for spent nuclear fuel or high level nuclear waste. This is generally thought to require fast reactors as most transuranic (TRU) isotopes have low fission probabilities in thermal reactors. Reduced-moderation LWRs are a potential alternative to fast reactors with reduced time to deployment as they are based on commercially mature LWR technology. Thorium (Th) fuel is neutronically advantageous for TRU multiple recycle in LWRs due to a large improvement in the void coefficient. If Th fuel is used in reduced-moderation LWRs, it appears neutronically feasible to achieve full actinide recycle while burning an external supply of TRU, with related potential improvements in waste management and fuel utilization. In this paper, the fuel cycle of TRU-bearing Th fuel is analysed for reduced-moderation PWRs and BWRs (RMPWRs and RBWRs). RMPWRs have the advantage of relatively rapid implementation and intrinsically low conversion ratios. However, it is challenging to simultaneously satisfy operational and fuel cycle constraints. An RBWR may potentially take longer to implement than an RMPWR due to more extensive changes from current BWR technology. However, the harder neutron spectrum can lead to favourable fuel cycle performance. A two-stage fuel cycle, where the first pass is Th-Pu MOX, is a technically reasonable implementation of either concept. The first stage of the fuel cycle can therefore be implemented at relatively low cost as a Pu disposal option, with a further policy option of full recycle in the medium term. (authors)

  2. METHUSELAH II - A Fortran program and nuclear data library for the physics assessment of liquid-moderated reactors

    International Nuclear Information System (INIS)

    Brinkworth, M.J.; Griffiths, J.A.

    1966-03-01

    METHUSELAH II is a Fortran program with a nuclear data library, used to calculate cell reactivity and burn-up in liquid-moderated reactors. It has been developed from METHUSELAH I by revising the nuclear data library, and by introducing into the program improvements relating to nuclear data, improvements in efficiency and accuracy, and additional facilities which include a neutron balance edit, specialised outputs, fuel cycling, and fuel costing. These developments are described and information is given on the coding and usage of versions of METHUSELAH II for the IBM 7030 (STRETCH), IBM 7090, and KDF9 computers. (author)

  3. Impact of moderator history on physics parameters in pressurized water reactors

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1988-01-01

    The magnitude of differential reactivity effects that result from spectral differences in different portions of a pressurized water (PWR) core is studied, and it is shown that these effects can be correlated very well with the local moderator history. The impact of these differences on physics parameters such as axial offset, isothermal moderator temperature coefficient, and differential control rod worth is shown to be significant for two PWRs of considerably different design

  4. Molten fuel-coolant interactions resulting from power transients in aluminium plate/water moderated reactors

    International Nuclear Information System (INIS)

    Storr, G.J.

    1989-08-01

    The behaviour of two reactors SL1 and SPERT D12, which underwent fast nuclear power transients prior to core destruction by a molten fuel-coolant interaction (MFCI) has been analysed and the results compared with measured data. The calculated spatial melt distribution and the mechanical work done during the events leads to high (∼ 250 kJ/kg) conversion efficiencies for this type of interaction when compared with molten drop experiments. A simple model for the steam explosion, using static thermodynamic properties of high temperature and pressure steam is used to calculate the dynamics of the reactors following the MFCI. 26 refs., 5 figs., 5 tabs

  5. Top Management Team Diversity and Company Performance: The moderating effect of Organization Life Cycle

    Directory of Open Access Journals (Sweden)

    Emil Velinov

    2016-12-01

    Full Text Available The research paper examines the moderating impact of Organizational Life Cycle on the relationship between Top Management Team Diversity and Company Performance. The study first elaborates and establishes the theoretical link between organization lifecycle and composition of management elites. Second, a quantitative empirical study is conducted to test the OLC stages moderating impact on the upper echelons diversity and firm performance of the top companies in the Czech Republic. A detailed procedure is developed to accurately classify organizations at different lifecycle stages, drawing extensively on existing literature and scales. Paper findings state that more mature the company becomes, more diversified the senior management is regardless the firm performance. Also, the industry dynamism impact has its own role in the relationship between the organization life cycle and senior management diversity which is expressed by the paper findings as well.

  6. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, R; Aalto, E

    1964-09-15

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

  7. Organic food consumption in Taiwan: Motives, involvement, and purchase intention under the moderating role of uncertainty.

    Science.gov (United States)

    Teng, Chih-Ching; Lu, Chi-Heng

    2016-10-01

    Despite the progressive development of the organic food sector in Taiwan, little is known about how consumers' consumption motives will influence organic food decision through various degrees of involvement and whether or not consumers with various degrees of uncertainty will vary in their intention to buy organic foods. The current study aims to examine the effect of consumption motives on behavioral intention related to organic food consumption under the mediating role of involvement as well as the moderating role of uncertainty. Research data were collected from organic food consumers in Taiwan via a questionnaire survey, eventually obtaining 457 valid questionnaires for analysis. This study tested the overall model fit and hypotheses through structural equation modeling method (SEM). The results show that consumer involvement significantly mediates the effects of health consciousness and ecological motives on organic food purchase intention, but not applied to food safety concern. Moreover, the moderating effect of uncertainty is statistical significance, indicating that the relationship between involvement and purchase intention becomes weaker in the condition of consumers with higher degree of uncertainty. Several implications and suggestions are also discussed for organic food providers and marketers. Copyright © 2016. Published by Elsevier Ltd.

  8. Study on characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2005-01-01

    Several characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states have been investigated. Performances of PWR and CANDU reactors are compared. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000 code. In the present study, we have compared the characteristics for different moderator to fuel ratio (MFR, 0.1 to 30), different burn-up for CANDU type and four fuels cycle schemes. Nuclide density of 235 U at MFR=1.9 decreases with increasing number of confined HM, while 235 U at higher MFR has the opposite trend. However, the nuclide density of fissile material at higher MFR is lower except 238 U. CANDU type requires lower uranium enrichment and obtains higher conversion ratio than PWR type. Lowest burn-up requires the lowest uranium enrichment and obtains the highest conversion ratio. The breeding condition can be obtained for plutonium recycle cases at MFR=2.1 of Case 4 and MFR=1.4 of Case 3. The natural uranium can be achieved at MFR=14 of plutonium recycle cases, and it can be used easier by increasing number of confined HM. (author)

  9. Reactivity and reaction rate ratio changes with moderator voidage in a light water high converter reactor lattice

    International Nuclear Information System (INIS)

    Chawla, R.; Gmuer, K.; Hager, H.; Seiler, R.

    1986-01-01

    Integral reaction rate ratios and other k ∞ related parameters have been measured in the first three cores of the experimental program on light water high converter reactor (LWHCR) test lattices in the PROTEUS reactor. The reference tight-pitch lattice consisted of two rod types, with an average fissile-plutonium enrichment of 6% and a fuel/moderator ratio of 2.0. The moderators were H 2 O, Dowtherm (simulating an H 2 O voidage of 42.5%), and air (100% void). Comparisons of the measured parameters have been made with calculational results based mainly on the use of two separate codes and their associated data libraries, namely, WIMS-D and EPRI-CPM. A reconstruction of individual components of the k-infinity void coefficient has been carried out on the basis of the measured changes with voidage of the various reaction rate ratios, as well as of k-infinity itself. The subsequent more detailed comparisons between experiment and calculation should provide a useful basis for resolving the conflicting calculational results that have been reported in the past for the void coefficient characteristics of LWHCRs. (author)

  10. Activity Of The Moderate And Radical Islamic Organizations In The European States

    Directory of Open Access Journals (Sweden)

    Nina V. Volodina

    2014-12-01

    Full Text Available In the present article author examines various aspects of Islamic organizations functioning in modern European countries. In the world the dangerous situation due to the spread of radical Islamic views, this is directly related to safety arose. Author researches some aspects of Islamic organizations in Europe functioning and their division into "moderate" and "radical". Author notes that the concept of "moderate Islam" is actively used in the western countries. In the article it is emphasized that supporters of such division believe that it is the "moderate Islamists" may help deal with the terrorist threat. Author gives definition of "religious extremism" and analyzes activities of Islamic organizations in Germany and France, as well as other European countries at the present stage of their development. In the process of study author pays enough attention to the notion of "spreading radicalism within the law", analyze existing legal acts. In the article problems of financing extremist organizations is pointed out and the need to create a set of joint measures to counter extremism and terrorism by Russian Federation and European countries, further improvement of law-enforcement is identified.

  11. Problems of two-phase flows in water cooled and moderated reactors

    International Nuclear Information System (INIS)

    Syu, Yu.

    1984-01-01

    Heat exchange in two-phase flows of coolant in loss of coolant accidents in PWR and BWR reactors has been investigated. Three main stages of accident history are considered: blowdown, reflooding using emergency core cooling system and rewetting. Factors, determining the rate of coolant leakage and the rate of temperature increase in fuel cladding during blowdown, processes of vapour during reflooding and liquid priming by vapour during rewetting, are discussed

  12. Numerical simulation of moderator flow and temperature distributions in a CANDU reactor vessel

    International Nuclear Information System (INIS)

    Carlucci, L.N.

    1982-10-01

    This paper describes numerical predictions of the two-dimensional flow and temperature fields of an internally-heated liquid in a typical CANDU reactor vessel. Turbulence momentum and energy transport are simulated using the k-epsilon model. Both steady-state and transient results are discussed. The finite control volume analogues of the conservation equations are solved using a modified version of the TEACH code

  13. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  14. Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

    International Nuclear Information System (INIS)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

  15. On the Use of Fine Distributed Moderating Material to Enhance Feedback Coefficients in Fast Reactors

    International Nuclear Information System (INIS)

    Merk, B.

    2012-01-01

    Conclusions and outlook: • Use of moderating material to enhance feedback coefficients in SFR: • Creation of a new degree of freedom for SFR design; • Good opportunities for the compensation of the effects due to transmutation fuels; • No major influence on Am incinieration; • Major problem is the thermal stability; • Stability up to ~1300°C by use of YH

  16. Copper extraction from coarsely ground printed circuit boards using moderate thermophilic bacteria in a rotating-drum reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Michael L.M., E-mail: mitchel.marques@yahoo.com.br [Bio& Hydrometallurgy Laboratory, Department of Metallurgical and Materials Engineering, Universidade Federal de Ouro Preto, Campus Morro do Cruzeiro, Ouro Preto, MG 35400-000 (Brazil); Leão, Versiane A., E-mail: versiane@demet.em.ufop.br [Bio& Hydrometallurgy Laboratory, Department of Metallurgical and Materials Engineering, Universidade Federal de Ouro Preto, Campus Morro do Cruzeiro, Ouro Preto, MG 35400-000 (Brazil); Gomes, Otavio [Centre for Mineral Technology – CETEM, Av Pedro Calmon, 900, 21941-908 Rio de Janeiro (Brazil); Lambert, Fanny; Bastin, David; Gaydardzhiev, Stoyan [Mineral Processing and Recycling, University of Liege, SartTilman, 4000 Liege (Belgium)

    2015-07-15

    Highlights: • Copper bioleaching from PCB (20 mm) by moderate thermophiles was demonstrated. • Larger PCB sheets enable a cost reduction due to the elimination of fine grinding. • Crushing generated cracks in PCB increasing the copper extraction. • A pre-treatment step was necessary to remove the lacquer coating. • High copper extractions (85%) were possible with pulp density of up to 25.0 g/L. - Abstract: The current work reports on a new approach for copper bioleaching from Printed Circuit Board (PCB) by moderate thermophiles in a rotating-drum reactor. Initially leaching of PCB was carried out in shake flasks to assess the effects of particle size (−208 μm + 147 μm), ferrous iron concentration (1.25–10.0 g/L) and pH (1.5–2.5) on copper leaching using mesophile and moderate thermophile microorganisms. Only at a relatively low solid content (10.0 g/L) complete copper extraction was achieved from the particle size investigated. Conversely, high copper extractions were possible from coarse-ground PCB (20 mm-long) working with increased solids concentration (up to 25.0 g/L). Because there was as the faster leaching kinetics at 50 °C Sulfobacillus thermosulfidooxidans was selected for experiments in a rotating-drum reactor with the coarser-sized PCB sheets. Under optimal conditions, copper extraction reached 85%, in 8 days and microscopic observations by SEM–EDS of the on non-leached and leached material suggested that metal dissolution from the internal layers was restricted by the fact that metal surface was not entirely available and accessible for the solution in the case of the 20 mm-size sheets.

  17. Copper extraction from coarsely ground printed circuit boards using moderate thermophilic bacteria in a rotating-drum reactor

    International Nuclear Information System (INIS)

    Rodrigues, Michael L.M.; Leão, Versiane A.; Gomes, Otavio; Lambert, Fanny; Bastin, David; Gaydardzhiev, Stoyan

    2015-01-01

    Highlights: • Copper bioleaching from PCB (20 mm) by moderate thermophiles was demonstrated. • Larger PCB sheets enable a cost reduction due to the elimination of fine grinding. • Crushing generated cracks in PCB increasing the copper extraction. • A pre-treatment step was necessary to remove the lacquer coating. • High copper extractions (85%) were possible with pulp density of up to 25.0 g/L. - Abstract: The current work reports on a new approach for copper bioleaching from Printed Circuit Board (PCB) by moderate thermophiles in a rotating-drum reactor. Initially leaching of PCB was carried out in shake flasks to assess the effects of particle size (−208 μm + 147 μm), ferrous iron concentration (1.25–10.0 g/L) and pH (1.5–2.5) on copper leaching using mesophile and moderate thermophile microorganisms. Only at a relatively low solid content (10.0 g/L) complete copper extraction was achieved from the particle size investigated. Conversely, high copper extractions were possible from coarse-ground PCB (20 mm-long) working with increased solids concentration (up to 25.0 g/L). Because there was as the faster leaching kinetics at 50 °C Sulfobacillus thermosulfidooxidans was selected for experiments in a rotating-drum reactor with the coarser-sized PCB sheets. Under optimal conditions, copper extraction reached 85%, in 8 days and microscopic observations by SEM–EDS of the on non-leached and leached material suggested that metal dissolution from the internal layers was restricted by the fact that metal surface was not entirely available and accessible for the solution in the case of the 20 mm-size sheets

  18. Study of burned optimization for minor actinides in European Sodium Fast Reactor (ESFR) by use of moderator materials

    International Nuclear Information System (INIS)

    Ramos, R L; Villanueva, A J; Buiront, L

    2012-01-01

    The minor actinides (MA) burn up optimization in the European Sodium Fast Reactor (ESFR) core was studied by adding different moderating materials in the Minor Actinides Bearing Blanket subassemblies (MABB SA) using the ERANOS neutron code package. These SA are of hexagonal shape and are composed of pellets inside of pins. These pellets contain a mixture of uranium dioxide (UO 2 ) and americium dioxide (AmO 2 ). If some of these pins are replaced by other identical ones containing moderating material instead of minor actinides, a shift in the spectrum towards lower energies is expected, which might enhance the burn up performance. The results of this work demonstrated that the use of compounds of hydrogen and magnesium as moderators produces a shift in the neutron spectrum, improving the porcentual minor actinides consumption. ZrH 2 moderator material was found to exhibit the best performances for this propose, followed by MgO and MgAl 2 O 4 , in that order. The use of SiC, BeO, TiC, LiO 2 and ZrC material produced no effect on the shift of the neutron spectrum. For safety reasons, it seems hardly realistic to use hydrogenous compounds in sodium fast reactors. So, compounds with magnesium are selected to be placed into the pins to improve the porcentual minor actinides consumption. The ESFR core is composed by 817 SA, 453 of them are fuel SA, 247 are reflectors SA, 84 are MABB (Minor Actinides Bearing Blankets) SA and 33 are control and shutdown rods. When about half of the total pins in each MABB were substituted by moderator pins with MgO pellets (135 of 271 pins), the porcentual consumption of minor actinides was of 30.85 %, i.e., 227.22 kg of minor actinides were consumed out of 736.65 kg in the initial configuration. In the case where all the pins of the MABB contained pellets of minor actinides, the porcentual consumption of minor actinides was of 21.26 %, i.e., 312.13 kg of minor actinides were consumed of 1467.87 kg in the initial configuration (author)

  19. 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

  20. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  1. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  2. Moderators for the design of a cold neutron source for the RA 3 reactor

    International Nuclear Information System (INIS)

    Cantargi, F; Sbaffoni, M; Granada, R

    2004-01-01

    The cold neutron production of hydrogenous materials was studied, taking into account their radiation resistance, for the conceptual design of a cold neutron source for the RA-3 reactor.Low spontaneous release of chemical energy was found in mesitylene.Libraries for hidrogen in mesitylene were generated using the NJOY nuclear processing system and the resulting cross sections were compared with experimental data.Good agreement between measurements and calculations was found in those cases where data are available.New calculations using the RA-3 geometry and these validated libraries will be performed [es

  3. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    Science.gov (United States)

    Gorman, Phillip Michael

    The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

  4. Detailed description of an SSAC at the facility level for light water moderated (off-load refueled) power reactor facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-03-01

    This report is intended to provide the technical details of an effective State Systems of Accounting for and Control of Nuclear Material (SSAC) which Member States may use, if they wish, to establish and maintain their SSACs. It is expected that systems designed along the lines described would be effective in meeting the objectives of both national and international systems for nuclear material accounting and control. This document accordingly provides a detailed description of a system for the accounting for and control of nuclear material in an off-load refueled light water moderated power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  5. Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

    International Nuclear Information System (INIS)

    Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

    2002-01-01

    A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

  6. Harnessing demographic differences in organizations: What moderates the effects of workplace diversity?

    Science.gov (United States)

    Dawson, Jeremy F.; Otaye‐Ebede, Lilian; Woods, Stephen A.; West, Michael A.

    2015-01-01

    Summary To account for the double‐edged nature of demographic workplace diversity (i.e,. relational demography, work group diversity, and organizational diversity) effects on social integration, performance, and well‐being‐related variables, research has moved away from simple main effect approaches and started examining variables that moderate these effects. While there is no shortage of primary studies of the conditions under which diversity leads to positive or negative outcomes, it remains unclear which contingency factors make it work. Using the Categorization‐Elaboration Model as our theoretical lens, we review variables moderating the effects of workplace diversity on social integration, performance, and well‐being outcomes, focusing on factors that organizations and managers have control over (i.e., strategy, unit design, human resource, leadership, climate/culture, and individual differences). We point out avenues for future research and conclude with practical implications. © 2015 The Authors. Journal of Organizational Behavior published by John Wiley & Sons, Ltd PMID:28239234

  7. Harnessing demographic differences in organizations: What moderates the effects of workplace diversity?

    Science.gov (United States)

    Guillaume, Yves R F; Dawson, Jeremy F; Otaye-Ebede, Lilian; Woods, Stephen A; West, Michael A

    2017-02-01

    To account for the double-edged nature of demographic workplace diversity (i.e,. relational demography, work group diversity, and organizational diversity) effects on social integration, performance, and well-being-related variables, research has moved away from simple main effect approaches and started examining variables that moderate these effects. While there is no shortage of primary studies of the conditions under which diversity leads to positive or negative outcomes, it remains unclear which contingency factors make it work. Using the Categorization-Elaboration Model as our theoretical lens, we review variables moderating the effects of workplace diversity on social integration, performance, and well-being outcomes, focusing on factors that organizations and managers have control over (i.e., strategy, unit design, human resource, leadership, climate/culture, and individual differences). We point out avenues for future research and conclude with practical implications. © 2015 The Authors. Journal of Organizational Behavior published by John Wiley & Sons, Ltd.

  8. Improvements in gas supply systems for heavy-water moderated reactors; Etudes de perfectionnements aux systemes d'alimentation en gaz d'un reacteur modere a l'eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, G; Hassig, J M; Laurent, N; Thomas, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [French] Dans un reacteur modere a l'eau lourde et refroidi au gaz sous pression, un probleme important du point de vue du trace du bloc pile et de son economie est le choix du systeme d'alimentation en gaz. Pour une solution a tubes de force, l'ensemble des structures du bloc reacteur est a temperature relativement faible, alors que les organes d'alimentation en gaz sont a celle, notablement plus elevee, du gaz. Ces organes, traverses par le debit du caloporteur, doivent lui opposer le minimum de resistance afin de ne pas necessiter un supplement onereux de puissance de

  9. Radial heat transfer from fuel to moderator during LOCAs for CANDU PHW reactors

    International Nuclear Information System (INIS)

    Hildebrandt, J.G.; So, C.B.; Gillespie, G.E.; MacLean, G.

    1983-01-01

    In a postulated CANDU-PHW loss-of-coolant accident (LOCA) with coincident impaired emergency cooling, the axial transport of heat from the fuel by convection is reduced. This reduction in heat removal causes the fuel to heat up and the radial heat transfer to the moderator to become significant. This paper deals with two codes that predict the thermal response of fuel channels under LOCA conditions. New channel thermal radiation models in both RAMA, a thermalhydraulic code, and CHAN II, a fuel channel thermo-chemical code, are presented and their predictions are compared with the experimental results of an electrically heated bundle of 37 fuel pins. A second experiment, involving a single heated pin in a channel with flowing steam, is presented. The predictions of RAMA and CHAN II are compared with this experiment to verify the codes' thermo-chemical models. There is good agreement between the predictions of both codes and the experimental results

  10. Technologies for tritium control in fission reactors moderated with heavy water

    International Nuclear Information System (INIS)

    Ramilo, L.B.; Gomez de Soler, S.M.

    1996-01-01

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors' belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the 'on-line' detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs

  11. Effective cross section values for well-moderated thermal reactor spectra

    Energy Technology Data Exchange (ETDEWEB)

    Westcott, C.H

    1970-11-15

    This document replaces the earlier versions (CRRP-680 and CRRP-787) and now employs the {sigma} (E) information contained in Supplement 1 (1960) of the 2nd edition of BNL-325 and other data privately collected to a cut-off date of April 1, 1960. The compilation is also enlarged to include higher temperatures (thus superseding CRRP-862) and for the first time also includes s-factors calculated using the epithermal cut-off functions exhibiting a maximum at just above the cut-off energy, which have recently been indicated by Swedish and U.K. measurements, as well as by analysis of some calculated spectra. As in the earlier compilation, the notation of the author's Geneva Conference (1958) paper is used, the effective cross section {sigma} being given in terms of the 2200 m/sec. value {sigma}{sub o} by the relation {sigma} {sigma}{sub o} (g + rs), where g and s are the factors listed in this compilation and r is a measure of the proportion of epithermal neutrons in the reactor spectrum. (author)

  12. Effective cross section values for well-moderated thermal reactor spectra

    Energy Technology Data Exchange (ETDEWEB)

    Westcott, C H

    1970-11-15

    This document replaces the earlier versions (CRRP-680 and CRRP-787) and now employs the {sigma} (E) information contained in Supplement 1 (1960) of the 2nd edition of BNL-325 and other data privately collected to a cut-off date of April 1, 1960. The compilation is also enlarged to include higher temperatures (thus superseding CRRP-862) and for the first time also includes s-factors calculated using the epithermal cut-off functions exhibiting a maximum at just above the cut-off energy, which have recently been indicated by Swedish and U.K. measurements, as well as by analysis of some calculated spectra. As in the earlier compilation, the notation of the author's Geneva Conference (1958) paper is used, the effective cross section {sigma} being given in terms of the 2200 m/sec. value {sigma}{sub o} by the relation {sigma} {sigma}{sub o} (g + rs), where g and s are the factors listed in this compilation and r is a measure of the proportion of epithermal neutrons in the reactor spectrum. (author)

  13. Effective cross section values for well-moderated thermal reactor spectra

    International Nuclear Information System (INIS)

    Westcott, C.H.

    1970-11-01

    This document replaces the earlier versions (CRRP-680 and CRRP-787) and now employs the σ (E) information contained in Supplement 1 (1960) of the 2nd edition of BNL-325 and other data privately collected to a cut-off date of April 1, 1960. The compilation is also enlarged to include higher temperatures (thus superseding CRRP-862) and for the first time also includes s-factors calculated using the epithermal cut-off functions exhibiting a maximum at just above the cut-off energy, which have recently been indicated by Swedish and U.K. measurements, as well as by analysis of some calculated spectra. As in the earlier compilation, the notation of the author's Geneva Conference (1958) paper is used, the effective cross section σ being given in terms of the 2200 m/sec. value σ o by the relation σ σ o (g + rs), where g and s are the factors listed in this compilation and r is a measure of the proportion of epithermal neutrons in the reactor spectrum. (author)

  14. Moderator Chemistry Program

    International Nuclear Information System (INIS)

    Dewitt, L.V.; Gibbs, A.; Lambert, D.P.; Bohrer, S.R.; Fanning, R.L.; Houston, M.W.; Stinson, S.L.; Deible, R.W.; Abdel-Khalik, S.I.

    1990-11-01

    Over the past fifteen months, the Systems Chemistry Group of the Reactor Engineering Department has undertaken a comprehensive study of the Department's moderator chemistry program at Savannah River Site (SRS). An internal review was developed to formalize and document this program. Objectives were as outlined in a mission statement and action plan. In addition to the mission statement and action plan, nine separate task reports have been issued during the course of this study. Each of these task reports is included in this document as a chapter. This document is an organized compilation of the individual reports issued by the Systems Chemistry Group in assessment of SRS moderator chemistry to determine if there were significant gaps in the program as ft existed in October, 1989. While these reviews found no significant gaps in that mode of operation, or any items that adversely affected safety, items were identified that could be improved. Many of the items have already been dear with or are in the process of completion under this Moderator Chemistry Program and other Reactor Restart programs. A complete list of the items of improvement found under this assessment is found in Chapter 9, along with a proposed time table for correcting remaining items that can be improved for the chemistry program of SRS reactors. An additional external review of the moderator chemistry processes, recommendations, and responses to/from the Reactor Corrosion Mitigation Committee is included as Appendix to this compilation

  15. Decommissioning the Romanian Water-Cooled Water-Moderated Research Reactor: New Environmental Perspective on the Management of Radioactive Waste

    International Nuclear Information System (INIS)

    Barariu, G.; Giumanca, R.

    2006-01-01

    Pre-feasibility and feasibility studies were performed for decommissioning of the water-cooled water-moderated research reactor (WWER) located in Bucharest - Magurele, Romania. Using these studies as a starting point, the preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as for the rehabilitation of the existing Radioactive Waste Treatment Plant and for the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and ecological reconstruction of the grounds need to be provided for, in accordance with national and international regulations. In accordance with IAEA recommendations at the time, the pre-feasibility study proposed three stages of decommissioning. However, since then new ideas have surfaced with regard to decommissioning. Thus, taking into account the current IAEA ideology, the feasibility study proposes that decommissioning of the WWER be done in one stage to an unrestricted clearance level of the reactor building in an Immediate Dismantling option. Different options and the corresponding derived preferred option for waste management are discussed taking into account safety measures, but also considering technical, logistical and economic factors. For this purpose, possible types of waste created during each decommissioning stage are reviewed. An approximate inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the recommended international basic safety standards identified in the previous phase of the project. The existing Radioactive Waste Treatment Plant (RWTP) from the Horia Hulubei Institute for Nuclear Physics and Engineering (IFIN-HH), which has been in service with no significant upgrade since 1974, will need refurbishing due to deterioration, as well as upgrading in order to ensure the

  16. Pressure loadings of Soviet-designed VVER [Water-Cooled, Water-Moderated Energy Reactor] reactor release mitigation structures from large-break LOCAs

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Horak, W.C.

    1989-01-01

    Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs

  17. Scope of activities and organization of an interuniversity reactor institute

    International Nuclear Information System (INIS)

    de Bruin, M.

    1990-01-01

    The Reactor Instituut Delft was founded in 1958 and was at that time part of the Delft University of Technology. In 1969, the institute was converted into an interuniversity institute, owned and directed by the combined Dutch universities. Since 1987, the institute has again constituted part of the Delft University of Technology, still continuing its role as an interuniversity institute and with provisions in the organizational structure to secure this role. The major facility of the institute is the Hoger Onderwijs reactor, a 2-MW swimming pool reactor operated 24 h/day, 5 days/week. The reactor is used in neutron beam studies, reactor physics research, and for isotope production, neutron activation analysis (NAA), and commercial irradiation. The institute's 3-MeV Van de Graaff electron accelerator is mainly used for radiation chemistry. It can deliver subnanosecond high-current pulses and is provided with fast optical and conductivity measuring equipment. The variable energy positron source is being used for the study of defects at metal surfaces and interfaces. The experience obtained with this source is used in the development of a much stronger source as the basis of a positron microbeam in one of the reactor beam tubes

  18. Safety aspects of long term operation of water moderated reactors. Recommendations on the scope and content of programmes for safe long term operation. Final report of the extrabudgetary programme on safety aspects long term operation of water moderated reactors

    International Nuclear Information System (INIS)

    2007-07-01

    During the last two decades, the number of IAEA Member States giving high priority to continuing the operation of nuclear power plants beyond the time frame originally anticipated is increasing. This is related to the age of nuclear power plants connected to the grid worldwide. The IAEA started to develop guidance on the safety aspects of ageing management in the 1990s. Recognizing the development in a number of its Member States, the IAEA initiated this Extrabudgetary Programme on Safety Aspects of Long Term Operation of Water Moderated Reactors in 2003. The objective of the Programme was to establish recommendations on the scope and content of activities to ensure safe long term operation of water moderated reactors. The term long term operation is used to accommodate various approaches in Member States and is defined as operation beyond an initial time frame set forth in design, standards, licence, and/or regulations, that is justified by safety assessment, considering life limiting processes and features for systems, structures and components. The scope of the Programme included general long term operation framework, mechanical components and materials, electrical components and instrumentation and control, and structural components and structures. The scope of the Programme was limited to physical structures of the NPPs. Four working groups addressed the above indicated technical areas. The Programme steering committee provided coordination and guidance and served as a forum for the exchange of information. The Programme implementation relied on voluntary in kind and financial contributions from the Czech Republic, Hungary, Slovakia, Sweden, the United Kingdom and the USA as well as in kind contributions from Bulgaria, Finland, the Netherlands, the Russian Federation, Spain, the Ukraine, and the European Commission. This report summarizes the main results, conclusions and recommendations of this Programme and provides in the Appendices I-IV detailed

  19. 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.)

  20. The operating organization and the recruitment, training and qualification of personnel for research reactors. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide provides recommendations on meeting the requirements on the operating organization and on personnel for research reactors. It covers the typical operating organization for research reactor facilities; the recruitment process and qualification in terms of education, training and experience; programmes for initial and continuing training; the authorization process for those individuals having an immediate bearing on safety; and the processes for their requalification and reauthorization

  1. Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.

    1996-01-01

    At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U 3 Si 2 fuel is about 6.0 g U/cm 3 . The French Commissariat a l'Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L'Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm 3 and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm 3 . On the other hand, UN is the least reactive fuel because of the relatively large 14 N(n,p) cross section. For a fixed value of k eff , the required 235 U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO 2 dispersions are only useful for uranium densities below 5.0 g/cm 3 . In this density range, however, UO 2 is more reactive than U 3 Si 2

  2. Effect of horizontal flow on the cooling of the moderator brick in the advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    Ganesan, P.; He, S.; Hamad, F.; Gotts, J.

    2011-01-01

    The paper reports an investigation of the effect of the horizontal cross flow on the temperature of the moderator brick in UK Advanced Gas-cooled Reactor (AGR) using computational fluid dynamics (CFD) with a conjugate heat transfer model for the solid and fluid. The commercial software package of ANSYS Fluent is used for this purpose. The CFD model comprises the full axial length of one-half of a typical fuel channel (assuming symmetry) and part of neighbouring channels on either side. Two sets of simulations have been carried out, namely, one with cross flow and one without cross flow. The effect of cross flow has subsequently been derived by comparing the results from the two groups of simulations. The study shows that a small cross flow can have a significant effect on the cooling of the graphite brick, causing the peak temperature of the brick to reduce significantly. Two mechanisms are identified to be responsible for this. Firstly, the small cross flow causes a significant redistribution of the main axial downward flow and this leads to an enhancement of heat transfer in some of the small clearances, and an impairment in others although overall, the enhancement is dominant leading to a better cooling. Secondly, the cross flow makes effective use of the small clearances between the key/keyway connections which increases the effective heat transfer area, hence increasing the cooling. Under the conditions of no cross flow, these areas remain largely inactive in heat transfer. The study shows that the cooling of the moderator is significantly enhanced by the cross flow perpendicular to the main cooling flow. (author)

  3. Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

    International Nuclear Information System (INIS)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung

    2015-01-01

    The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW th power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H 2 corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the preliminary design

  4. Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW{sub th} power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H{sub 2} corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the

  5. Effects of core models and neutron energy group structures on xenon oscillation in large graphite-moderated reactors

    International Nuclear Information System (INIS)

    Yamasita, Kiyonobu; Harada, Hiroo; Murata, Isao; Shindo, Ryuichi; Tsuruoka, Takuya.

    1993-01-01

    Xenon oscillations of large graphite-moderated reactors have been analyzed by a multi-group diffusion code with two- and three-dimensional core models to study the effects of the geometric core models and the neutron energy group structures on the evaluation of the Xe oscillation behavior. The study clarified the following. It is important for accurate Xe oscillation simulations to use the neutron energy group structure that describes well the large change in the absorption cross section of Xe in the thermal energy range of 0.1∼0.65 eV, because the energy structure in this energy range has significant influences on the amplitude and the period of oscillations in power distributions. Two-dimensional R-Z models can be used instead of three-dimensional R-θ-Z models for evaluation of the threshold power of Xe oscillation, but two-dimensional R-θ models cannot be used for evaluation of the threshold power. Although the threshold power evaluated with the R-θ-Z models coincides with that of the R-Z models, it does not coincide with that of the R-θ models. (author)

  6. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  7. Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    International Nuclear Information System (INIS)

    Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B.; Fong, R.W.L.; Coleman, C.E.

    1998-08-01

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHIF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 degrees C. (author)

  8. 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

  9. Fast Reactor Knowledge Organization System (FR-KOS)

    International Nuclear Information System (INIS)

    Costantini, Livio

    2013-01-01

    FR-KOS Web-portal general concepts: FR-KOS web-portal is an integrated and unique solution for knowledge preservation, retrieval and management in fast reactor (FR) knowledge domain; The exploitation of the FR-KOS provides an effective and flexible tool to access FR knowledge repositories. The dynamic and interactive process of searching information is guided by structures that are based on the FR Taxonomy; FR-KOS can be extended to other types of reactor and technologies, such as PWR, WWER, BWR, CANDU, etc. In addition can also be utilized in other information management process such as the preservation of nuclear accidents knowledg;. FR-KOS addresses needs in non-structured knowledge preservation for both embarking countries and countries with established nuclear power; FR-KOS repositories contain more that 50.000 samples of technical and scientific publications, including full text documents in different languages (English, German, French and eventually in Russian)

  10. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    International Nuclear Information System (INIS)

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

    2013-01-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  11. Hydrolysis and degradation of filtrated organic particulates in a biofilm reactor under anoxic and aerobic conditions

    DEFF Research Database (Denmark)

    Janning, K.F.; Mesterton, K.; Harremoës, P.

    1997-01-01

    Two experiments were performed in order to investigate the anoxic and the aerobic degradation of filtrated organic matter in a biofilter. In submerged lab: scale reactors with Biocarbone media as filter material, accumulated particulate organic matter from pre-settled wastewater served as the only...

  12. 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

  13. Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

    Science.gov (United States)

    Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

    2018-01-04

    A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

  14. A study concerning tritium concentration evolution in the moderator of a CANDU reactor connected on-line to a detritiation facility

    International Nuclear Information System (INIS)

    Bidica, Nicolae; Bornea, Anisia

    2005-01-01

    The present work is a theoretical study on the tritium concentration evolution in the CANDU reactor moderator connected on-line with a detritiation facility. This study is based on a calculation model which takes into account the evolution curve of the tritium concentration in the absence of detritiation process in both the moderator and SPTC of the Unit 1 CANDU reactor at Cernavoda NPP. This study leads to determination of the tritium concentration evolution in the moderator in the presence of the detritiation process for both a range of intake flows and initial concentration. Also, the intake flow change will be analyzed for a detritiation facility as a function of tritium initial concentration existing in the moderator in the case of a survey of the detritiation over a given period of time. The conclusions of this study were the following: - an optimum of the detritiation factor can be determined; - detritiation starts at a lower value for the tritium concentration in moderator which reduces the strain upon the detritiation facility and therefore the costs of its building, maintenance and operation. (authors)

  15. Development of a noise-based method for the determination of the moderator temperature coefficient of reactivity (MTC) in pressurized water reactors (PWRs)

    International Nuclear Information System (INIS)

    Demaziere, C.

    2002-01-01

    The Moderator Temperature Coefficient of reactivity (MTC) is an important safety parameter of Pressurized Water Reactors (PWRs). In most countries, the so-called at-power MTC has to be measured a few months before the reactor outage, in order to determine if the MTC will not become too negative. Usually, the at-power MTC is determined by inducing a change in the moderator temperature, which has to be compensated for by other means, such as a change in the boron concentration. An MTC measurement using the boron dilution method is analysed in this thesis. It is demonstrated that the uncertainty of such a measurement technique is so large, that the measured MTC could become more negative than what the Technical Specifications allow. Furthermore, this technique incurs a disturbance of the plant operation. For this reason, another technique relying on noise analysis was proposed a few years ago. In this technique, the MTC is inferred from the neutron noise measured inside the core and the moderator temperature noise measured at the core-exit, in the same or in a neighbouring fuel assembly. This technique does not require any perturbation of the reactor operation, but was nevertheless proven to underestimate the MTC by a factor of 2 to 5. In this thesis, it is shown, both theoretically and experimentally, that the reason of the MTC underestimation by noise analysis is the radially loosely coupled character of the moderator temperature noise throughout the core. A new MTC noise estimator, accounting for this radially non-homogeneous moderator temperature noise is proposed and demonstrated to give the correct MTC value. This new MTC noise estimator relies on the neutron noise measured in a single point of the reactor and the radially averaged moderator temperature noise measured inside the core. In the case of the Ringhals-2 PWR in Sweden, Gamma-Thermometers (GTs) offer such a possibility since in dynamic mode they measure the moderator temperature noise, whereas in static

  16. 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.

  17. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors

    International Nuclear Information System (INIS)

    Martin Munoz, M.; Alvarez Gonzalez, F.

    1969-01-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs

  18. Stability of an anaerobic single reactor filled with dolomitic limestone with increased organic load of sugarcane

    Directory of Open Access Journals (Sweden)

    Maria Magdalena Ribas Döll

    2017-12-01

    Full Text Available The anaerobic single-stage reactor was evaluated to treat vinasse and to evaluate its stability. This bench reactor was filled with dolomitic limestone with a horizontal plug flow to simulate a drainage channel. The experiment lasted 129 days while the reactor was submitted to different applied organic concentrations (chronologically applied: 3.0; 5.0; 12.0; 9.0 and 7.5 g L-1 as COD, chemical oxygen demand. COD removals were 50% and 9% with 3.0 and 7.5 g L-1, respectively. With 12.0 g L-1, reactor efficiency increased to 33%, with an abrupt drop to 3% on the 84th day. Therefore, in order to avoid reactor collapse, a remedial measure was necessary. The system remained in batch without feeding for 19 days (from the 85th to the 104th day with 9.0 g L-1. Afterwards, it was observed that the performance of the system tended to stabilize, reaching 47% with 7.5 g L-1 in the 118th day. At the end of the experiment, the potassium content of the wastewater decreased from 800 mg L-1 to 594 mg L-1 (on an average 25% and calcium and magnesium increased within the reactor liquor. The dissolution of the limestone inside the liquor reactor probably caused this result. After the treatment with limestone, the average pH value of the effluent increased from 4.9 to over 6.0 in all organic concentrations. It could be concluded that the reactor filled with dolomitic limestone in these operational conditions assured a low efficiency in COD removal, potassium reduction, increasing values of pH, alkalinity, calcium and magnesium. The instability was observed when there was increase in organic load to 12 g L-1 with subsequent recovery.

  19. Behavior of radioactive organic iodide in an atmosphere of High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Saeki, Masakatsu; Nakashima, Mikio; Sagawa, Chiaki; Masaki, Nobuyuki; Hirabayashi, Takakuni; Aratono, Yasuyuki

    1990-06-01

    Formation and decomposition behavior of radioactive organic iodide have been studied in an atmosphere of High Temperature Gas-cooled Reactor (High Temperature Engineering Test Reactor, HTTR). Na 125 I was chosen for radioactive iodine source instead of CsI diffusing from coated fuel particles. Na 125 I adsorbed on graphite was heated in pure He and He containing O 2 or H 2 O atmosphere. The results obtained are as follows. It was proved that organic iodide was formed with organic radicals released from graphite even in He atmosphere. Thus, the interchange rate of inorganic iodide with organic iodide was remarkably decreased with prolonged preheat-treatment period at 1000degC. Organic iodide formed was easily decomposed by its recirculation into hot reaction tube kept at 900degC. When organic iodide was passed through powdered graphite bed, more than 70% was decomposed at 90degC. Oxygen and water vapour intermixed in He suppressed the interchange rate of inorganic iodide with organic iodide. These results suggest that organic iodide rarely exists in the pressure vessel under normal operating condition of HTTR, and, under hypothetical accident condition of HTTR, organic iodide fraction never exceeds the value used for a safety assessment of light water reactor. (author)

  20. Microbial Population Dynamics and Ecosystem Functions of Anoxic/Aerobic Granular Sludge in Sequencing Batch Reactors Operated at Different Organic Loading Rates

    Directory of Open Access Journals (Sweden)

    Enikö Szabó

    2017-05-01

    Full Text Available The granular sludge process is an effective, low-footprint alternative to conventional activated sludge wastewater treatment. The architecture of the microbial granules allows the co-existence of different functional groups, e.g., nitrifying and denitrifying communities, which permits compact reactor design. However, little is known about the factors influencing community assembly in granular sludge, such as the effects of reactor operation strategies and influent wastewater composition. Here, we analyze the development of the microbiomes in parallel laboratory-scale anoxic/aerobic granular sludge reactors operated at low (0.9 kg m-3d-1, moderate (1.9 kg m-3d-1 and high (3.7 kg m-3d-1 organic loading rates (OLRs and the same ammonium loading rate (0.2 kg NH4-N m-3d-1 for 84 days. Complete removal of organic carbon and ammonium was achieved in all three reactors after start-up, while the nitrogen removal (denitrification efficiency increased with the OLR: 0% at low, 38% at moderate, and 66% at high loading rate. The bacterial communities at different loading rates diverged rapidly after start-up and showed less than 50% similarity after 6 days, and below 40% similarity after 84 days. The three reactor microbiomes were dominated by different genera (mainly Meganema, Thauera, Paracoccus, and Zoogloea, but these genera have similar ecosystem functions of EPS production, denitrification and polyhydroxyalkanoate (PHA storage. Many less abundant but persistent taxa were also detected within these functional groups. The bacterial communities were functionally redundant irrespective of the loading rate applied. At steady-state reactor operation, the identity of the core community members was rather stable, but their relative abundances changed considerably over time. Furthermore, nitrifying bacteria were low in relative abundance and diversity in all reactors, despite their large contribution to nitrogen turnover. The results suggest that the OLR has

  1. The Effect of Organic Loading Rate on Milk WastewaterTreatment Using Sequencing Batch Reactor (SBR

    Directory of Open Access Journals (Sweden)

    Hooman Hajiabadi

    2009-09-01

    Full Text Available In this study, four aerobic sequencing batch reactors (SBRs were operated under the same conditions for the treatment of milk wastewater at different organic loading rates (OLRs. Cylindrical Plexiglas reactors were run for 56 days (including 21 days of acclimatization and 35 days of data gathering. Effective volume, influent wastewater flowrate, and sludge retention time (SRT of reactors were 5.5 L, 3.5 L/d, and 10 d, respectively. The average COD removal efficiency for the reactors R1, R2, R3, and R4 with influent OLRave values of 633, 929, 1915, and 3261 gCOD/m3d were 95, 96, 95, and 82 percent, respectively. The average effluent suspended solid (SS for all reactors was lower than 44 mg/L. Also, except for R4 with an average effluent turbidity of 270 NTU, other reactors met the Iranian wastewater emission standard (50 NTU. In addition, the average sludge volume index of reactors R1 to R3 was found to be lower than 67 mL/g. According to the results, the overall variation of COD removal efficiency versus influent OLR shows a decreasing rate with a correlation factor of 0.8 (R2.

  2. Space reactor/organic Rankine conversion - A near-term state-of-the-art solution

    Science.gov (United States)

    Niggemann, R. E.; Lacey, D.

    The use of demonstrated reactor technology with organic Rankine cycle (ORC) power conversion can provide a low cost, minimal risk approach to reactor-powered electrical generation systems in the near term. Several reactor technologies, including zirconium hydride, EBR-II and LMFBR, have demonstrated long life and suitability for space application at the operating temperature required by an efficient ORC engine. While this approach would not replace the high temperature space reactor systems presently under development, it could be available in a nearer time frame at a low and predictable cost, allowing some missions requiring high power levels to be flown prior to the availability of advanced systems with lower specific mass. Although this system has relatively high efficiency, the heat rejection temperature is low, requiring a large radiator on the order of 3.4 sq m/kWe. Therefore, a deployable heat pipe radiator configuration will be required.

  3. 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)

  4. 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.)

  5. Mockup tests of void fraction in moderator cell and two-phase thermosiphon loop of cold neutron source in China Advanced Research Reactor

    International Nuclear Information System (INIS)

    Du Shejiao; Bi Qincheng; Chen Tingkuan; Feng Quanke; Li Xiaoming

    2004-01-01

    Full-scale mockup tests were carried out using freon-113 as a working fluid to verify the design of China Advanced Research Reactor (CARR) Cold neutron Source (CNS), which is a two-phase hydrogen thermosiphon loop consisting of an annular cylindrical moderator cell, two separated hydrogen transfer tubes and a condenser. The circulation characteristics, liquid level and void fraction in the moderator cell against the variation of the heat load were studied. The density ratio and the volumetric evaporating rate of the mockup test are kept the same as those of CARR CNS. The test results show that the mockup loop can establish stable circulation and has a self-regulating characteristic. Within the moderator cell, the inner shell contains only vapor and the outer shell contains the mixture of vapor-liquid with void fraction in a certain range. (authors)

  6. Target organ specific activity of drosophila MRP (ABCC1) moderates developmental toxicity of methylmercury.

    Science.gov (United States)

    Prince, Lisa; Korbas, Malgorzata; Davidson, Philip; Broberg, Karin; Rand, Matthew Dearborn

    2014-08-01

    Methylmercury (MeHg) is a ubiquitous and persistent neurotoxin that poses a risk to human health. Although the mechanisms of MeHg toxicity are not fully understood, factors that contribute to susceptibility are even less well known. Studies of human gene polymorphisms have identified a potential role for the multidrug resistance-like protein (MRP/ABCC) family, ATP-dependent transporters, in MeHg susceptibility. MRP transporters have been shown to be important for MeHg excretion in adult mouse models, but their role in moderating MeHg toxicity during development has not been explored. We therefore investigated effects of manipulating expression levels of MRP using a Drosophila development assay. Drosophila MRP (dMRP) is homologous to human MRP1-4 (ABCC1-4), sharing 50% identity and 67% similarity with MRP1. A greater susceptibility to MeHg is seen in dMRP mutant flies, demonstrated by reduced rates of eclosion on MeHg-containing food. Furthermore, targeted knockdown of dMRP expression using GAL4>UAS RNAi methods demonstrates a tissue-specific function for dMRP in gut, Malpighian tubules, and the nervous system in moderating developmental susceptibility to MeHg. Using X-ray synchrotron fluorescence imaging, these same tissues were also identified as the highest Hg-accumulating tissues in fly larvae. Moreover, higher levels of Hg are seen in dMRP mutant larvae compared with a control strain fed an equivalent dose of MeHg. In sum, these data demonstrate that dMRP expression, both globally and within Hg-targeted organs, has a profound effect on susceptibility to MeHg in developing flies. Our findings point to a potentially novel and specific role for dMRP in neurons in the protection against MeHg. Finally, this experimental system provides a tractable model to evaluate human polymorphic variants of MRP and other gene variants relevant to genetic studies of mercury-exposed populations. © The Author 2014. Published by Oxford University Press on behalf of the Society of

  7. An organic profile of a pressurised water reactor secondary plant

    International Nuclear Information System (INIS)

    Eeden, Nestor van; Stwayi, Mandisibuntu; Gericke, Gerhard

    2012-01-01

    Make-up water addition to the steam/water cycle at Koeberg Nuclear Power Station usually results in a corresponding increase of the chloride concentration in the steam generator blowdown system. During plant transients, when higher than normal make-up is required to the secondary plant, the concentration of chloride occasionally exceeds the limiting value for the station chemistry performance indicator. Irrespective of this, the demineralised water make-up supply tanks, which are routinely analysed for chloride, are within all recognised acceptable standards for secondary water make-up and therefore these tanks do not initially appear to be the source of chloride contamination. Water treatment at the plant relies essentially on ion exchange, which has been proven to be very effective in removing inorganic ionic species such as chloride. Organic compounds are less effectively removed by ion exchange and may pass through the treatment system, and these organics can reside undetected in the make-up water tanks. Historically, the elevated chloride concentration following high system make-up has been attributed to chlorinated organic compounds known as trihalomethanes being present in the make-up water tanks, but no rigorous study had been undertaken. As it has been assumed that the majority of chloride in the secondary system originates from the make-up water organic impurities, it was considered important to confirm this by compiling an organic profile of the secondary plant. The use of organic additives was also taken into account in the profile. This work has confirmed the contribution from trihalomethanes and has also found that other organochlorides contribute even more significantly to the overall chloride inventory of the secondary plant. (orig.)

  8. 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.)

  9. In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

    International Nuclear Information System (INIS)

    Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

    2000-01-01

    During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

  10. Dynamic model of organic pollutant degradation in three dimensional packed bed electrode reactor.

    Science.gov (United States)

    Pang, Tianting; Wang, Yan; Yang, Hui; Wang, Tianlei; Cai, Wangfeng

    2018-04-21

    A dynamic model of semi-batch three-dimensional electrode reactor was established based on the limiting current density, Faraday's law, mass balance and a series of assumptions. Semi-batch experiments of phenol degradation were carried out in a three-dimensional electrode reactor packed with activated carbon under different conditions to verify the model. The factors such as the current density, the electrolyte concentration, the initial pH value, the flow rate of organic and the initial organic concentration were examined to know about the pollutant degradation in the three-dimensional electrode reactor. The various concentrations and logarithm of concentration of phenol with time were compared with the dynamic model. It was shown that the calculated data were in good agreement with experimental data in most cases. Copyright © 2018 Elsevier Ltd. All rights reserved.

  11. Selected power reactor projects in Canada and the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-11-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power.

  12. Selected power reactor projects in Canada and the United States of America

    International Nuclear Information System (INIS)

    1964-01-01

    As part of its activities in connection with the development of nuclear power, the IAEA has undertaken a continuing study of the technology and economics of power reactors, with particular reference to the needs of the developing countries. Information on the progress made in eight power reactor projects, namely those of Bonus, Pathfinder, Elk River, Piqua, Hallam, Experimental Gas-Cooled Reactor (EGCR), High-Temperature Gas-Cooled Reactor (HTGCR) and Nuclear Power Demonstration (NPD), is presented in this report. Developments during the past year are shown, emphasis being placed on operating experience in the case of those reactors which have become critical. The Agency is grateful to the Governments of Canada and the USA, who have extended the necessary facilities for covering he different power reactor projects in their respective countries. The cooperation received from the reactor manufacturers, builders and operators is also gratefully acknowledged. It is hoped that this report will be of interest to reactor technologists and operators and those interested in the application of nuclear power

  13. Comparison of secondary organic aerosol formation from toluene on initially wet and dry ammonium sulfate particles at moderate relative humidity

    Directory of Open Access Journals (Sweden)

    T. Liu

    2018-04-01

    Full Text Available The formation of secondary organic aerosol (SOA has been widely studied in the presence of dry seed particles at low relative humidity (RH. At higher RH, initially dry seed particles can exist as wet particles due to water uptake by the seeds as well as the SOA. Here, we investigated the formation of SOA from the photooxidation of toluene using an oxidation flow reactor in the absence of NOx under a range of OH exposures on initially wet or dry ammonium sulfate (AS seed particles at an RH of 68 %. The ratio of the SOA yield on wet AS seeds to that on dry AS seeds, the relative SOA yield, decreased from 1.31 ± 0.02 at an OH exposure of 4.66 × 1010 molecules cm−3 s to 1.01 ± 0.01 at an OH exposure of 5.28 × 1011 molecules cm−3 s. This decrease may be due to the early deliquescence of initially dry AS seeds after being coated by highly oxidized toluene-derived SOA. SOA formation lowered the deliquescence RH of AS and resulted in the uptake of water by both AS and SOA. Hence the initially dry AS seeds contained aerosol liquid water (ALW soon after SOA formed, and the SOA yield and ALW approached those of the initially wet AS seeds as OH exposure and ALW increased, especially at high OH exposure. However, a higher oxidation state of the SOA on initially wet AS seeds than that on dry AS seeds was observed at all levels of OH exposure. The difference in mass fractions of m ∕ z 29, 43 and 44 of SOA mass spectra, obtained using an aerosol mass spectrometer (AMS, indicated that SOA formed on initially wet seeds may be enriched in earlier-generation products containing carbonyl functional groups at low OH exposures and later-generation products containing acidic functional groups at high exposures. Our results suggest that inorganic dry seeds become at least partially deliquesced particles during SOA formation and hence that ALW is inevitably involved in the SOA formation at moderate RH. More laboratory

  14. Comparison of secondary organic aerosol formation from toluene on initially wet and dry ammonium sulfate particles at moderate relative humidity

    Science.gov (United States)

    Liu, Tengyu; Huang, Dan Dan; Li, Zijun; Liu, Qianyun; Chan, ManNin; Chan, Chak K.

    2018-04-01

    The formation of secondary organic aerosol (SOA) has been widely studied in the presence of dry seed particles at low relative humidity (RH). At higher RH, initially dry seed particles can exist as wet particles due to water uptake by the seeds as well as the SOA. Here, we investigated the formation of SOA from the photooxidation of toluene using an oxidation flow reactor in the absence of NOx under a range of OH exposures on initially wet or dry ammonium sulfate (AS) seed particles at an RH of 68 %. The ratio of the SOA yield on wet AS seeds to that on dry AS seeds, the relative SOA yield, decreased from 1.31 ± 0.02 at an OH exposure of 4.66 × 1010 molecules cm-3 s to 1.01 ± 0.01 at an OH exposure of 5.28 × 1011 molecules cm-3 s. This decrease may be due to the early deliquescence of initially dry AS seeds after being coated by highly oxidized toluene-derived SOA. SOA formation lowered the deliquescence RH of AS and resulted in the uptake of water by both AS and SOA. Hence the initially dry AS seeds contained aerosol liquid water (ALW) soon after SOA formed, and the SOA yield and ALW approached those of the initially wet AS seeds as OH exposure and ALW increased, especially at high OH exposure. However, a higher oxidation state of the SOA on initially wet AS seeds than that on dry AS seeds was observed at all levels of OH exposure. The difference in mass fractions of m / z 29, 43 and 44 of SOA mass spectra, obtained using an aerosol mass spectrometer (AMS), indicated that SOA formed on initially wet seeds may be enriched in earlier-generation products containing carbonyl functional groups at low OH exposures and later-generation products containing acidic functional groups at high exposures. Our results suggest that inorganic dry seeds become at least partially deliquesced particles during SOA formation and hence that ALW is inevitably involved in the SOA formation at moderate RH. More laboratory experiments conducted with a wide variety of SOA precursors

  15. Anaerobic digestion of grain stillage at high organic loading rates in three different reactor systems

    International Nuclear Information System (INIS)

    Schmidt, Thomas; Pröter, Jürgen; Scholwin, Frank; Nelles, Michael

    2013-01-01

    In this study the anaerobic digestion of grain stillage in three different reactor systems (continuous stirred tank reactor, anaerobic sequencing batch reactor, fixed bed reactor) with and without immobilization of microorganisms was investigated to evaluate the performance during increase of the organic loading rate (OLR) from 1 to 10 g of volatile solids (VS) per liter reactor volume and day and decrease of the hydraulic retention time (HRT) from 40 to 6 days. No significant differences have been observed between the performances of the three examined reactor systems. The changes in OLR and HRT caused a reduction of the specific biogas production (SBP) of about 25% from about 650 to 550 L kg −1 of VS but would also diminish the necessary digester volume and investment costs of about 75% compared to the state of the art. -- Highlights: ► It was shown that without immobilization of microorganisms low HRT's are possible. ► No significant differences have been observed between different digester designs. ► Trace element supplementation is obligatory with grain stillage as substrate

  16. Cell-Culture Reactor Having a Porous Organic Polymer Membrane

    Science.gov (United States)

    Koontz, Steven L. (Inventor)

    2000-01-01

    A method for making a biocompatible polymer article using a uniform atomic oxygen treatment is disclosed. The substrate may be subsequently optionally grated with a compatibilizing compound. Compatibilizing compounds may include proteins, phosphory1choline groups, platelet adhesion preventing polymers, albumin adhesion promoters, and the like. The compatibilized substrate may also have a living cell layer adhered thereto. The atomic oxygen is preferably produced by a flowing afterglow microwave discharge, wherein the substrate resides in a sidearm out of the plasma. Also, methods for culturing cells for various purposes using the various membranes are disclosed as well. Also disclosed are porous organic polymers having a distributed pore chemistry (DPC) comprising hydrophilic and hydrophobic regions, and a method for making the DPC by exposing the polymer to atomic oxygen wherein the rate of hydrophilization is greater than the rate of mass loss.

  17. Kinetic characterization for hemicellulose hydrolysis of corn stover in a dilute acid cycle spray flow-through reactor at moderate conditions

    International Nuclear Information System (INIS)

    Jin, Qiang; Zhang, Hongman; Yan, Lishi; Qu, Liang; Huang, He

    2011-01-01

    The kinetic characterization of hemicellulose hydrolysis of corn stover was investigated using a new reactor of dilute acid cycle spray flow-through (DCF) pretreatment. The primary purpose was to obtain kinetic data for hemicellulose hydrolysis with sulfuric acid concentrations (10-30 kg m -3 ) at relatively low temperatures (90-100 o C). A simplified kinetic model was used to describe its performance at moderate conditions. The results indicate that the rates of xylose formation and degradation are sensitive to flow rate, temperature and acid concentration. Moreover, the kinetic data of hemicellulose hydrolysis fit a first-order reaction model and the experimental data with actual acid concentration after accounting for the neutralization effect of the substrates at different temperatures. Over 90% of the xylose monomer yield and below 5.5% of degradation product (furfural) yield were observed in this reactor. Kinetic constants for hemicellulose hydrolysis models were analyzed by an Arrhenius-type equation, and the activation energy of xylose formation were 111.6 kJ mol -1 , and 95.7 kJ mol -1 for xylose degradation, respectively. -- Highlights: → Investigating a novel pretreatment reactor of dilute acid cycle spray flow-through. → Xylose yield is sensitive to flow rate, temperature and acid concentration. → Obtaining relatively higher xylose monomer yield and lower fermentation inhibitor. → Lumping hemicellulose and xylan oligmers together in the model is a valid way. → The kinetic model as a guide for reactor design, and operation strategy optimization.

  18. CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

    2002-10-01

    In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

  19. Outline of design, manufacturing and installation experience of pressure vessel structure for the prototype heavy water moderated boiling light water cooled reactor 'FUGEN'

    International Nuclear Information System (INIS)

    Shibato, Eizo; Oguchi, Isao; Kishi, Toshikazu; Kitagawa, Yuji

    1977-01-01

    After component installation completed in June 1977 and various functional tests to be conducted later, the prototype heavy water moderated, boiling light water cooled reactor ''FUGEN'' is scheduled to reach first criticality in March 1978. Since the pressure vessel of ''FUGEN'' is completely different from that of a light water reactor in structure and materials, through research and development work was carried out prior to fabrication and construction. Based on these studies, installation of the actual pressure vessel was completed. Functional tests are now under way. This article describes examples in which our research and development results are reflected on design, manufacture, and installation of the pressure vessel. Also it introduces noteworthy achievements relevant to production techniques in manufacture and installation. (auth.)

  20. Development of quantitative analytical procedures on two-phase flow in tight-lattice fuel bundles for reduced-moderation light-water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Takae, K.; Tamai, H.; Akimoto, H.; Yoshida, H.

    2004-01-01

    The research project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR is a light water reactor for which a higher conversion ratio more than one can be expected. In order to attain this higher conversion ratio, triangular tight-lattice fuel bundles whose gap spacing between each fuel rod is around 1 mm are required. As for the thermal design of the RMWR core, conventional analytical methods are no good because the conventional composition equations can not predict the RMWR core with high accuracy. Then, development of new quantitative analytical procedures was carried out. Those analytical procedures are constructed by model experiments and advanced two-phase flow analysis codes. This paper describes the results of the model experiments and analytical results with the developed analysis codes. (authors)

  1. Transient classification for the IRIS reactor using self-organized maps built in free platform

    International Nuclear Information System (INIS)

    Doraskevicius Junior, Waldemar

    2005-01-01

    Kohonen's Self Organized Maps (SOM) were tested with data from several operational conditions of the nuclear reactor IRIS (International Reactor Innovative and Secure) to develop an effective tool in the classification and transient identification in nuclear reactors. The data were derived from 56 simulations of the operation of IRIS, from steady-state conditions to accidents. The digital system built for the tests was based on the JAVA platform for the portability and scalability, and for being one of the free development platforms. Satisfactory results of operation classification were obtained with reasonable processing time in personal computers; about two to five minutes were spent for ordination and convergence of the learning on the data base. The methodology of this work was extended to the supervision of logistics of natural gas for Brazilian pipelines, showing satisfactory results for the classification of deliveries for simultaneous measurement in several points. (author)

  2. Development of an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated reactors

    International Nuclear Information System (INIS)

    Babaev, N.S.

    1981-06-01

    The results of work carried out under IAEA Contract No. 2336/RB are described (subject: an automated system of nuclear materials accounting for nuclear power stations with water-cooled, water-moderated (VVER) reactors). The basic principles of an accounting system for this type of nuclear power plant are outlined. The general structure and individual units of the information computer program used to achieve automated accounting are described and instructions are given on the use of the program. A detailed example of its application (on a simulated nuclear power plant) is examined

  3. 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.

  4. 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.

  5. Melting of contaminated steel scrap from the dismantling of the CO2 systems of gas cooled, graphite moderated nuclear reactors

    International Nuclear Information System (INIS)

    Feaugas, J.; Jeanjacques, M.; Peulve, J.

    1994-01-01

    G2 and G3 are the natural Uranium cooled reactors Graphite/Gas. The two reactors were designed for both plutonium and electricity production (45 MWe). The dismantling of the reactors at stage 2 has produced more than 4 000 tonnes of contaminated scrap. Because of their large mass and low residual contamination level, the French Atomic Energy Commission (CEA) considered various possibilities for the processing of these metallic products in order to reduce the volume of waste going to be stored. After different studies and tests of several processes and the evaluation of their results, the choice to melt the dismantled pipeworks was taken. It was decided to build the Nuclear Steel Melting Facility known as INFANTE, in cooperation with a steelmaker (AHL). The realization time schedule for the INFANTE lasted 20 months. It included studies, construction and the licensing procedure. (authors). 2 tabs., 3 figs

  6. 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.

  7. Design of a mixed recharge with MOX assemblies of greater relation of moderation for a BWR reactor; Diseno de una recarga mixta con ensambles MOX de mayor relacion de moderacion para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G.; Palacios H, J. [ININ, Carretera Mexico-Toluca Km. 36.5, 52045 Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2004-07-01

    The study of the fuel of mixed oxides of uranium and plutonium (MOX) it has been topic of investigation in many countries of the world and those are even discussed in many places the benefits of reprocessing the spent fuel to extract the plutonium created during the irradiation of the fuel in the nuclear power reactors. At the moment those reactors that have been loaded partially with MOX fuel, are mainly of the type PWR where a mature technology has been achieved in some countries like they are France, Belgium and England, however the experience with reactors of the type BWR is more limited and it is continued studying the best way to introduce this type of fuel in BWRs, one of the main problems to introduce MOX in reactors BWR is the neutronic design of the same one, existing different concepts to introduce the plutonium in the assemblies of fuel and one of them is the one of increasing the relationship of moderation of the assemble. In this work a MOX fuel assemble design is presented and the obtained results so far in the ININ. These results indicate that the investigated concept has some exploitable advantages in the use of the MOX fuel. (Author)

  8. Early Attachment Organization Moderates the Parent-Child Mutually Coercive Pathway to Children's Antisocial Conduct

    Science.gov (United States)

    Kochanska, Grazyna; Barry, Robin A.; Stellern, Sarah A.; O'Bleness, Jessica J.

    2009-01-01

    This multimethod study of 101 mothers, fathers, and children elucidates poorly understood role of children's attachment security as "moderating" a common maladaptive trajectory: from parental power assertion, to child resentful opposition, to child antisocial conduct. Children's security was assessed at 15 months, parents' power assertion observed…

  9. Organization of the ITER [International Thermonuclear Experimental Reactor] Project - Sharing of information and procurements

    International Nuclear Information System (INIS)

    Shannon, T.E.

    1990-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is expected to fully confirm the scientific feasibility and to address the technological feasibility of fusion power. Consequently, the machine must be designed for controlled ignition and extended burn of deuterium-tritium plasma. It must also demonstrate and perform integrated testing of components required to utilize fusion power for practical purposes. Cooperation among four countries/organizations (United States, Soviet Union, Japan, and EURATOM) to build a single experimental reactor will reduce the cost for each country and provide an international pool of scientific and engineering resources. This paper describes ITER organization for conceptual design activity, schedule for conceptual design activities, ITER operating parameters, conceptual project schedule and cost, future plans, basic principles and problems related to task sharing, and basic principles in handling of intellectual property

  10. Organic insulators and the copper stabilizer for fusion-reactor magnets

    International Nuclear Information System (INIS)

    Coltman, R.R. Jr.

    1981-11-01

    The materials which compose the large composite superconducting fusion reactor magnets are subjected to mechanical stress, neutron and gamma-ray radiation with broad energy spectra, high magnetic fields, and thermal cycling from 4 to 300 K. Of the materials now considered for use in the magnets, results show that the organic insulators and the Cu stabilizer are the most sensitive to this environment. In response to the need for stabilizer data, magnetoresistivity changes were studied in eight variously prepared specimens of Cu throughout five cycles of an alternate neutron irradiation (4.0 K) and annealing (14 h at 307 K) program. The results were combined with those on the radiation behavior of epoxy and polyimide organic insulators to provide a preliminary assessment of their comparative radiation resistance in a typical magnet location of the Experimental Power Reactor

  11. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

    The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

  12. Organization and mechanization of maintenance operations at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    Titov, A.A.

    1983-01-01

    The structure of capital investments defining organization and mechanization of maintepance operations at NPPs with the WWER type reactors is analyzed. The trends in development of optimum decisions for organization and mechanization of repair obs at NPPs being designed taking into account the prospects of nuclear powep enginerning development, the system of NPP maintenance servicing, as well as the structure of repair-productive capacities are discussed. On the basis of the analysis of the data obtained in designing the Zaporozhskaya NPP it is shown that the capital investments for organizing and mechanization of maintenance operations at the unified NPP site with four WWER-1000 reactors reach nearly 18 roubles/kW. A conclusion is drawn that at present the design of an NPP with the WWER-1000 reactor totally meets the requirements of realization of periodic maintenance operations. It is advisable to cooperate the NPP management with that of a thermal power station from the viewpoint of using manpower, which would improve the operating conditions and labour productivity of workers engaged in repair and, consequently, reduce the capital investments and repair expenditures

  13. Secondary organic aerosol from VOC mixtures in an oxidation flow reactor

    Science.gov (United States)

    Ahlberg, Erik; Falk, John; Eriksson, Axel; Holst, Thomas; Brune, William H.; Kristensson, Adam; Roldin, Pontus; Svenningsson, Birgitta

    2017-07-01

    The atmospheric organic aerosol is a tremendously complex system in terms of chemical content. Models generally treat the mixtures as ideal, something which has been questioned owing to model-measurement discrepancies. We used an oxidation flow reactor to produce secondary organic aerosol (SOA) mixtures containing oxidation products of biogenic (α-pinene, myrcene and isoprene) and anthropogenic (m-xylene) volatile organic compounds (VOCs). The resulting volume concentration and chemical composition was measured using a scanning mobility particle sizer (SMPS) and a high-resolution time-of-flight aerosol mass spectrometer (HR-ToF-AMS), respectively. The SOA mass yield of the mixtures was compared to a partitioning model constructed from single VOC experiments. The single VOC SOA mass yields with no wall-loss correction applied are comparable to previous experiments. In the mixtures containing myrcene a higher yield than expected was produced. We attribute this to an increased condensation sink, arising from myrcene producing a significantly higher number of nucleation particles compared to the other precursors. Isoprene did not produce much mass in single VOC experiments but contributed to the mass of the mixtures. The effect of high concentrations of isoprene on the OH exposure was found to be small, even at OH reactivities that previously have been reported to significantly suppress OH exposures in oxidation flow reactors. Furthermore, isoprene shifted the particle size distribution of mixtures towards larger sizes, which could be due to a change in oxidant dynamics inside the reactor.

  14. Harnessing demographic differences in organizations: What moderates the effects of workplace diversity?

    OpenAIRE

    Guillaume, YRF; Dawson, JF; Otaye-Ebede, L; Woods, SA; West, MA

    2015-01-01

    To account for the double-edged nature of demographic workplace diversity (i.e. relational demography, work group diversity, and organizational diversity) effects on social integration, performance and well-being related variables, research has moved away from simple main effect approaches and started examining variables that moderate these effects. While there is no shortage of primary studies of the conditions under which diversity leads to positive or negative outcomes, it remains unclear ...

  15. Operation of a catalytic reverse flow reactor for the purification of air contamined with volatile organic compounds

    NARCIS (Netherlands)

    van de Beld, L.; van de Beld, L.; Westerterp, K.R.

    1997-01-01

    Catalytic oxidation in a reverse flow reactor is an attractive process for the decontamination of air polluted with volatile organic compounds (VOCs). In this paper several aspects of operating this type of reactor for air purification under strongly varying conditions will be discussed. For a

  16. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors; Determinacion espectrografica de impurezas metalicas en refrigerantes organicos para reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Martin Munoz, M; Alvarez Gonzalez, F

    1969-07-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs.

  17. Modified kinetic-hydraulic UASB reactor model for treatment of wastewater containing biodegradable organic substrates.

    Science.gov (United States)

    El-Seddik, Mostafa M; Galal, Mona M; Radwan, A G; Abdel-Halim, Hisham S

    2016-01-01

    This paper addresses a modified kinetic-hydraulic model for up-flow anaerobic sludge blanket (UASB) reactor aimed to treat wastewater of biodegradable organic substrates as acetic acid based on Van der Meer model incorporated with biological granules inclusion. This dynamic model illustrates the biomass kinetic reaction rate for both direct and indirect growth of microorganisms coupled with the amount of biogas produced by methanogenic bacteria in bed and blanket zones of reactor. Moreover, the pH value required for substrate degradation at the peak specific growth rate of bacteria is discussed for Andrews' kinetics. The sensitivity analyses of biomass concentration with respect to fraction of volume of reactor occupied by granules and up-flow velocity are also demonstrated. Furthermore, the modified mass balance equations of reactor are applied during steady state using Newton Raphson technique to obtain a suitable degree of freedom for the modified model matching with the measured results of UASB Sanhour wastewater treatment plant in Fayoum, Egypt.

  18. Reaction Cross Section Calculations in Neutron Induced Reactions and GEANT4 Simulation of Hadronic Interactions for the Reactor Moderator Material BeO

    Directory of Open Access Journals (Sweden)

    Veli ÇAPALI

    2016-05-01

    Full Text Available BeO is one of the most common moderator material for neutron moderation; due to its high density, neutron capture cross section and physical-chemical properties that provides usage at elevated temperatures. As it’s known, for various applications in the field of reactor design and neutron capture, reaction cross–section data are required. The cross–sections of (n,α, (n,2n, (n,t, (n,EL and (n,TOT reactions for 9Be and 16O nuclei have been calculated by using TALYS 1.6 Two Component Exciton model and EMPIRE 3.2 Exciton model in this study. Hadronic interactions of low energetic neutrons and generated isotopes–particles have been investigated for a situation in which BeO was used as a neutron moderator by using GEANT4, which is a powerful simulation software. In addition, energy deposition along BeO material has been obtained. Results from performed calculations were compared with the experimental nuclear reaction data exist in EXFOR.

  19. 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...

  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. Special Analysis: Atmospheric Dose Resulting from the Release of C14 from Reactor Moderator Deionizers in a Disposal Environment

    International Nuclear Information System (INIS)

    Hiergesell, Robert A.; Swingle, Robert F.

    2005-01-01

    The proposed action of disposing of 52 moderator deionizer vessels within the ILV was evaluated in this SA. In particular, a detailed analysis of the release of 14 C via the atmospheric pathway was conducted for these vessels since the major concern has been the nearly 20 Ci of 14 C that is associated with each vessel. The more rigorous evaluation of the atmospheric pathway for 14 C included incorporation of new information about the chemical availability of 14 C when disposed in a grout/cement encapsulation environment, as will be the case in the ILV. This information was utilized to establish the source term for a 1-D numerical model to simulate the diffusion of 14 CO 2 from the ILV Waste Zone to the land surface. The results indicate a peak surface emanation rate from the entire ILV of 1.42E-08 Ci/yr with an associated dose of only 3.83E-05 mrem/yr to the Maximally Exposed Individual (MEI) at 100m. The fact that the atmospheric pathway exposure for 14 C is controlled by chemical solubility limits for 14 C between the solid waste, pore water and pore vapor within the disposal environment rather than the absolute inventory suggests that the establishment of specific facility limits is inappropriate. With the relaxation of the atmospheric pathway restriction, the groundwater pathway becomes the more restrictive in terms of disposing 14 C or 14 C KB within the ILV. Since the resin-based 14 C of the 52 moderator deionizer vessels is highly similar to the 14 C KB waste form, the inventory from the 52 deionizer vessels is compared against the groundwater limits for that waste form. The small groundwater pathway fraction (1.14E-05) calculated for the proposed inventory of the 52 moderator deionizer vessels indicates that the proposed action will have an insignificant impact with respect to possible exposures via the groundwater pathway. This investigation recommends that there be no ILV Atmospheric pathway limit for 14 C and 14 C KB . Further, in the absence of an

  2. Reflector drums as control mechanism for craft thermionic reactors with constant emitter heating containing U-233 as fuel and beryllium as moderator

    International Nuclear Information System (INIS)

    Sahin, S.; Selvi, S.

    1980-01-01

    The suitability of borated reflector drums has been investigated and shown as a control mechanism for space craft thermionic reactors with constant emitter heating using U-233 as fuel and beryllium to be moderator, mainly due to their extremce compactness and their very soft neutron sepctrum. The achievable change in ksub(eff) allows long-term control operation with success. The use of reflector drums keeps the cone diameter and the mass of the radiation shield on minimum. The distortion of the emitter heating field remains under acceptable tolerances, mainly due to the enhanced neutron production at the outer core region and the remaining reflector part between the boron layer and the core. All neutron physics calculations have been carried out using the multigroup Ssub(N) methods. Three data groups for r-theta-calculations in S 4 -P 1 approximation (16 space angles) have been evaluated from a 123-energy-groups data library using transport theoretical methods. (orig.) [de

  3. Technologies for tritium control in fission reactors moderated with heavy water; Tecnologias para control de tritio en reactores de fision moderados con agua pesada

    Energy Technology Data Exchange (ETDEWEB)

    Ramilo, L B; Gomez de Soler, S M [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    This study was done within a program one of whose objectives was to analyze the possible strategies and technologies, to be applied to HWR at Argentine nuclear power plants, for tritium control. The high contribution of tritium to the total dose has given rise to the need by the operators and/or designers to carry out developments and improvements to try to optimize tritium control technologies. Within a tritium control program, only that one which includes the heavy water detritiation will allow to reduce the tritium concentrations at optimum levels for safety and cost-effective power plant operation. The technology chosen to be applied should depend not only on the technical feasibility but also on the analysis of economic and juncture factors such as, among others, the quantity of heavy water to be treated. It is the authors` belief that AECL tendency concerning heavy water treatment in its future reactors would be to employ the CECE technology complemented with immobilization on titanium beds, with the `on-line` detritiation in each nuclear power plant. This would not be of immediate application since our analysis suggests that AECL would assume that the process is under development and needs to be tested. (author). 21 refs.

  4. Nuclear reactor core having nuclear fuel and composite burnable absorber arranged for power peaking and moderator temperature coefficient control

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1991-01-01

    This patent describes a nuclear reactor core. It comprises a first group of fuel rods containing fissionable material and being free of burnable absorber material; and a second group of fuel rods containing fissionable material and first and second burnable absorber material; the first burnable absorber material being a boron-bearing material which does not contain erbium and the second burnable absorber material being an erbium material; the first and second burnable absorber materials being in the form of an outer coating on the fissionable material, the outer coating being composed of an inner layer of one of the boron-bearing material which does not contain erbium and the erbium material and an outer layer of the other of the boron-bearing material which does not contain erbium and the erbium material

  5. 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.

  6. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Hareux, F.; Roche, R.; Vrillon, B.

    1964-01-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [fr

  7. Effects of the addition of an organic polymer on the hydrolysis of sodium tetrahydroborate in batch reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, M.J.F.; Pinto, A.M.F.R. [Centro de Estudos de Fenomenos de Transporte, Departamento de Engenharia Quimica, Faculdade de Engenharia da Universidade do Porto, Rua Dr. Roberto Frias s/n, 4200-465 Porto (Portugal); Fernandes, V.R.; Rangel, C.M. [Laboratorio Nacional de Energia e Geologia - LNEG, Fuel Cells and Hydrogen Unit, Estrada do Paco do Lumiar 22, 1649-038 Lisboa (Portugal); Gales, L. [Instituto de Biologia Molecular e Celular, Universidade do Porto, Rua do Campo Alegre 823, 4150-180 Porto and Instituto de Ciencias Biomedicas Abel Salazar, Largo Prof. Abel Salazar 2, 4099-003 Porto (Portugal)

    2010-10-15

    An experimental study is presented both on the generation and storage of molecular hydrogen (H{sub 2}) by small additions of an organic polymer -carboxymethyl cellulose (CMC) - to sodium borohydride (NaBH{sub 4}) through the alkaline hydrolysis, in the presence of a powdered nickel-ruthenium based catalyst reused from 274 to 282 times. The experiments were performed at 45 C in two batch reactors with internal volumes of 0.229 L and 0.369 L, made of stainless-steel with bottom conical shape, positioned vertically. The results showed that working at moderate pressures, up to 2.7 MPa, increases slightly the H{sub 2} dissolution in the liquid phase, enhanced by the changing of the polarity of the remained solution inside the reactor: a value of 0.182 for dimensionless H{sub 2} solubility in the liquid phase with 0.25 wt% CMC was found, at 45 C, based on Henry's law. As a consequence, sodium tetrahydroxoborate, NaB(OH){sub 4} by-product was produced in the presence of CMC additive, showing the absence of crystalline water in its crystal structure (NaB(OH){sub 4} presents structural water, with boron atoms linked to four hydroxyl groups). This new finding never reported to form at < 50 C, has a positive impact in recyclability costs of NaBO{sub 2} back to NaBH{sub 4} due to the elimination of two energy consuming steps in the metaborate dehydration kinetics. In fact our system of compressed hydrogen, shows that both H{sub 2} generation rates and yields and hydrogen storage capacities can be augmented, the latter to reach {approx} 6 wt%, by adding small amounts of an organic polymer (CMC) to the classic NaBH{sub 4} hydrolysis, performed with stoichiometric amount of water. The eventual success of this new route will depend upon developing a advantageous method of converting borates into tetrahydroborate and also finding materials (chemicals) which enhance the solubility of H{sub 2

  8. Business-nonprofit partnerships as a driver of internal marketing in nonprofit organizations. Consequences for nonprofit performance and moderators

    Directory of Open Access Journals (Sweden)

    Luis Ignacio Álvarez-González

    2017-04-01

    Full Text Available Nonprofit organizations (NPOs confront competitive pressures derived from complex economic and societal challenges. Their capacity to fulfil their mission increasingly depends on developing successful alliances with key external and internal stakeholders, including cooperative interorganizational relationships. In this context, the aim of this research is to analyze: (1 to which extent business-nonprofit partnerships (BNPPs foster the development of an internal marketing approach by NPOs; (2 the impact of this approach to human resource management on nonprofit performance; and (3 the possible moderating effect of the funding strategy of the nonprofit. This empirical research, based on a survey to a representative sample of Spanish NPOs, shows that cooperative relationships between nonprofit and business organizations are closely associated with a process of knowledge transfer, resulting in improved nonprofit performance; although these positive effects depend on the capacity of NPOs to generate income from commercial sources.

  9. How community organizations moderate the effect of armed conflict on migration in Nepal

    Science.gov (United States)

    Williams, Nathalie E.

    2013-01-01

    This article describes an analytical study of systematic micro-level variability in migration during conflict. The study is based on a multi-dimensional model of individual out-migration that examines the economic, social, and political consequences of conflict and how community organizations condition the experience of these consequences and systematically alter migration patterns. A unique combination of detailed data on violent events and individual behaviours during the Maoist insurrection in Nepal and multi-level event-history models were used to empirically test the model. Results indicate that community organizations dampened the effect of conflict on out-migration by providing resources that helped people to cope with the danger as well as economic, social and political consequences of conflict. This evidence suggests a systematic redistribution of population, partially contingent upon specific resources available in each community, which will likely affect the socio-demographic context of post-conflict Nepal into the future. PMID:23356735

  10. Membrane contactor/separator for an advanced ozone membrane reactor for treatment of recalcitrant organic pollutants in water

    International Nuclear Information System (INIS)

    Chan, Wai Kit; Jouët, Justine; Heng, Samuel; Yeung, King Lun; Schrotter, Jean-Christophe

    2012-01-01

    An advanced ozone membrane reactor that synergistically combines membrane distributor for ozone gas, membrane contactor for pollutant adsorption and reaction, and membrane separator for clean water production is described. The membrane reactor represents an order of magnitude improvement over traditional semibatch reactor design and is capable of complete conversion of recalcitrant endocrine disrupting compounds (EDCs) in water at less than three minutes residence time. Coating the membrane contactor with alumina and hydrotalcite (Mg/Al=3) adsorbs and traps the organics in the reaction zone resulting in 30% increase of total organic carbon (TOC) removal. Large surface area coating that diffuses surface charges from adsorbed polar organic molecules is preferred as it reduces membrane polarization that is detrimental to separation. - Graphical abstract: Advanced ozone membrane reactor synergistically combines membrane distributor for ozone, membrane contactor for sorption and reaction and membrane separator for clean water production to achieve an order of magnitude enhancement in treatment performance compared to traditional ozone reactor. Highlights: ► Novel reactor using membranes for ozone distributor, reaction contactor and water separator. ► Designed to achieve an order of magnitude enhancement over traditional reactor. ► Al 2 O 3 and hydrotalcite coatings capture and trap pollutants giving additional 30% TOC removal. ► High surface area coating prevents polarization and improves membrane separation and life.

  11. Life role salience and subjective well-being among Macedonian employees: Does family-supportive organization perception moderate this relationship.

    Science.gov (United States)

    Blaževska Stoilkovska, Biljana; Šurbanovska, Orhideja; Fritzhand, Ana; Stojanoska Ivanova, Tatjana

    2018-01-15

    As many studies have shown, one of the most important tendencies of employees nowadays is to achieve work- life balance. Organizations should develop various activities and create supportive climate, within the framework of which employees will have opportunities to realize aforementioned goals which in turn would increase work productivity and work motivation. The aim of this paper was to examine how subjective well-being (life satisfaction and exhaustion) is associated with commitment to work and family roles under the conditions of strongly and weakly perceived organizational support for family life among health care professionals, teachers and bankers in Macedonia. Marriage duration and the number of children were introduced as control variables. This cross-sectional study was conducted on a sample of 198 full-time employed doctors, nurses, teachers and bankers. Research variables were assessed using self-reported measures/ questionnaires. Hierarchical multiple linear regression was performed for data analysis. It was revealed that occupational role commitment contributed to highly expressed life satisfaction, while exhaustion was predicted by marital role commitment. These relationships were stronger among surveyed employees who reported positive family-supportive organization perception, but tested moderation effect of this variable was not statistically significant. Findings demonstrated that family-supportive organization perception moderated association of the number of children with life satisfaction, that is, participants who perceived an organization as family-supportive and had more children were more satisfied with their life in general. Results highlighted the importance of organizational orientation toward employees, their commitment to work and family roles, and their subjective well-being, as characteristics that might contribute to higher work engagement, success and family satisfaction. Int J Occup Med Environ Health 2018;31(3):281-291. This

  12. 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.)

  13. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  14. Analysis of neutron spectra and fluxes obtained with cold and thermal moderators at IBR-2 reactor: experimental and computer modeling studies at small-angle scattering YuMO setup

    International Nuclear Information System (INIS)

    Kuklin, A.I.; Rogov, A.D.; Gorshkova, Yu.E.; Kovalev, Yu.S.; Kutuzov, S.A.; Utrobin, P.K.; Rogachev, A.V.; Ivan'kov, O.I.; Solov'ev, D.V.; Gordelij, V.I.

    2011-01-01

    Results of experimental and computer modeling investigations of neutron spectra and fluxes obtained with cold and thermal moderators at the IBR-2 reactor (JINR, Dubna) are presented. The studies are done for small-angle neutron scattering (SANS) spectrometer YuMO (beamline number 4 of the IBR-2). The measurements of neutron spectra for two methane cold moderators are done for the standard configuration of the SANS instrument. The data from both moderators under different conditions of their operation are compared. The ratio of experimentally determined neutron fluxes of cold and thermal moderators at different wavelength is shown. Monte Carlo simulations are done to determine spectra for cold methane and thermal moderators. The results of the calculations of the ratio of neutron fluxes of cold and thermal moderators at different wavelength are demonstrated. In addition, the absorption of neutrons in the air gaps on the way from the moderator to the investigated sample is presented. SANS with the protein apoferritin was done in the case of cold methane as well as a thermal moderator and the data were compared. The perspectives for the use of the cold moderator for a SANS spectrometer at the IBR-2 are discussed. The advantages of the YuMO spectrometer with the thermal moderator with respect to the tested cold moderator are shown

  15. Analysis of a gas-liquid film plasma reactor for organic compound oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Hsieh, Kevin [Department of Chemical and Biomedical Engineering, Florida State University, Tallahassee, FL 32310 (United States); Wang, Huijuan [School of Environmental and Safety Engineering, Jiangsu University, Zhenjiang 212013 (China); Locke, Bruce R., E-mail: blocke@fsu.edu [Department of Chemical and Biomedical Engineering, Florida State University, Tallahassee, FL 32310 (United States)

    2016-11-05

    Highlights: • Non-homogeneous filamentary plasma discharge formed along gas-liquid interface. • Hydrogen peroxide formed near interface favored over organic oxidation from liquid. • Post-plasma Fenton reactions lead to complete utilization of hydrogen peroxide. - Abstract: A pulsed electrical discharge plasma formed in a tubular reactor with flowing argon carrier gas and a liquid water film was analyzed using methylene blue as a liquid phase hydroxyl radical scavenger and simultaneous measurements of hydrogen peroxide formation. The effects of liquid flow rate, liquid conductivity, concentration of dye, and the addition of ferrous ion on dye decoloration and degradation were determined. Higher liquid flow rates and concentrations of dye resulted in less decoloration percentages and hydrogen peroxide formation due to initial liquid conductivity effects and lower residence times in the reactor. The highest decoloration energy yield of dye found in these studies was 5.2 g/kWh when using the higher liquid flow rate and adding the catalyst. The non-homogeneous nature of the plasma discharge favors the production of hydrogen peroxide in the plasma-liquid interface over the chemical oxidation of the organic in the bulk liquid phase and post-plasma reactions with the Fenton catalyst lead to complete utilization of the plasma-formed hydrogen peroxide.

  16. Experience in the recruitment, organization and training of operations and maintenance personnel for the Malaysian research reactor

    International Nuclear Information System (INIS)

    Jamal Khair Ibrahim.

    1983-01-01

    The TRIGA Reactor located at the Tun Ismail Atomic Research Centre (PUSPATI) Complex is owned and operated by the Nuclear Energy Unit of the Prime Minster's Department. The operations and maintenance personnel are part and parcel of the national civil service organization. As such, the requirement and remuneration of these personnel are handled by a central federal government personnel management agency in common with personnel from other federal government agencies. In addition, the reactor is the first and only one in Malaysia, a developing country, which is the process of committing herself towards a nuclear power programme. These factors coupled with the absence of an independent reactor operator licensing agency posed unique problems in the recruitment, organization, training and licensing of operations personnel for the facility. The paper discusses these factors and their bearing on the recruitment, training, licensing and career development prospects of the PUSPATI TRIGA Reactor operators. (author)

  17. Optical Properties of Moderately-Absorbing Organic and Mixed Organic/Inorganic Particles at Very High Humidities

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Tami C; Rood, Mark J; Brem, Benjamin T; Mena-Gonzalez, Francisco C; Chen, Yanju

    2012-04-16

    Relative humidity (RH) affects the water content of an aerosol, altering its ability to scatter and absorb light, which is important for aerosol effects on climate and visibility. This project involves in situ measurement and modeling of aerosol optical properties including absorption, scattering and extinction at three visible wavelengths (467, 530, 660 nm), for organic carbon (OC) generated by pyrolysis of biomass, ammonium sulfate and sodium chloride, and their mixtures at controlled RH conditions. Novel components of this project include investigation of: (1) Changes in all three of these optical properties at scanned RH conditions; (2) Optical properties at RH values up to 95%, which are usually extrapolated instead of measured; and (3) Examination of aerosols generated by the pyrolysis of wood, which is representative of primary atmospheric organic carbon, and its mixture with inorganic aerosol. Scattering and extinction values were used to determine light absorption by difference and single scattering albedo values. Extensive instrumentation development and benchmarking with independently measured and modeled values were used to obtain and evaluate these new results. The single scattering albedo value for a dry absorbing polystyrene microsphere benchmark agreed within 0.02 (absolute value) with independently published results at 530 nm. Light absorption by a nigrosin (sample light-absorbing) benchmark increased by a factor of 1.24 +/-0.06 at all wavelengths as RH increased from 38 to 95%. Closure modeling with Mie theory was able to reproduce this increase with the linear volume average (LVA) refractive index mixing rule for this water soluble compound. Absorption by biomass OC aerosol increased by a factor of 2.1 +/- 0.7 and 2.3 +/- 1.2 between 32 and 95% RH at 467 nm and 530 nm, but there was no detectable absorption at 660 nm. Additionally, the spectral dependence of absorption by OC that was observed with filter measurements was confirmed qualitatively

  18. Effect of two heavy metals, cadmium and nickel, on the organic load removal efficiency in a laboratory UASB reactor

    International Nuclear Information System (INIS)

    Forero, Luis Eduardo; Sierra, Jorge Humberto

    2004-01-01

    Experiments were carried out in three up flow anaerobic sludge blanket, UASB, reactors each with 3 L capacity, four hours of hydraulic retention time, (HRT) and volumetric organic load of 4,8 g/L/d. After the initial start phase, which was of 4.000 hours for the three reactors, they were affected in the following way: the first reactor was continuously feed with 5 mg/L of cadmium chloride, the second one was continuously feed with 10 mg/L of nickel chloride and the last one was not affected and served as reference. Efficiency in organic load removal was measured as oxygen chemical demand (OCD), the first reactor changed from 60% in the start phase (phase one) to 18% in the cadmium-affected phase (phase two), efficiency in removal (OCI) in reactor two varied from 60 to 24% and the last one did not change in a noticeable manner. Reactor one accumulated cadmium in the mud, whereas reactor two did not do that with nickel

  19. Evaluation of organic endocrine disruptors in water at Brazilian Multipurpose Reactor - RMB installation area

    International Nuclear Information System (INIS)

    Silva, Larissa L.; Martins, Elâine A.J.; Cotrim, Marycel E.B.; Pires, Maria A.F.

    2017-01-01

    The study of pollutants with organic and inorganic characteristics in groundwater and surface waters of a given region is an important tool in the assessment of pollution. Endocrine interferers are synthetic or natural chemicals that have the ability to act on the endocrine system of humans and animals by mimetizing natural hormones and may produce adverse effects on organisms, even in low concentrations (μg or ng.L -1 ). Anthropic activities are the major source of input of endocrine disruptors into the environment. The Brazilian government has a project to construct a multipurpose reactor, Brazilian Multipurpose Reactor (RMB), at the Iperó city, to improve the nuclear research Brazilian capacity. The object of this research in to analyze 14 organic compounds that may be present in the groundwater and surface waters of the RMB installation area. This is an unprecedented and extremely important study for the evaluated region; since it will provide guidance on the degree of contamination of the local waters before the construction begins. The study will also make it possible to verify if the construction of the RMB will offer environmental issues to the place. For the determination of the compounds of interest, a developed and validated analytical method was used. This methodology consists of the concentration of the samples by solid phase extraction (SPE) followed by quantification by gas chromatography coupled to the mass spectrometry detector (GC/MS). The water column particulate was also evaluated by ultrasonic extraction, followed by quantification by GC/MS. The results reveal that some of the compounds were found and it was due to anthropic activities in the vicinity of the regions. By initial analysis it was possible verify river that cross the RMB area present values below 0.05 μg L -1 . (author)

  20. Evaluation of organic endocrine disruptors in water at Brazilian Multipurpose Reactor - RMB installation area

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Larissa L.; Martins, Elâine A.J.; Cotrim, Marycel E.B.; Pires, Maria A.F., E-mail: larissa.limeira07@gmail.com, E-mail: elaine@ipen.br, E-mail: mecotrim@ipen.br, E-mail: mapires@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The study of pollutants with organic and inorganic characteristics in groundwater and surface waters of a given region is an important tool in the assessment of pollution. Endocrine interferers are synthetic or natural chemicals that have the ability to act on the endocrine system of humans and animals by mimetizing natural hormones and may produce adverse effects on organisms, even in low concentrations (μg or ng.L{sup -1}). Anthropic activities are the major source of input of endocrine disruptors into the environment. The Brazilian government has a project to construct a multipurpose reactor, Brazilian Multipurpose Reactor (RMB), at the Iperó city, to improve the nuclear research Brazilian capacity. The object of this research in to analyze 14 organic compounds that may be present in the groundwater and surface waters of the RMB installation area. This is an unprecedented and extremely important study for the evaluated region; since it will provide guidance on the degree of contamination of the local waters before the construction begins. The study will also make it possible to verify if the construction of the RMB will offer environmental issues to the place. For the determination of the compounds of interest, a developed and validated analytical method was used. This methodology consists of the concentration of the samples by solid phase extraction (SPE) followed by quantification by gas chromatography coupled to the mass spectrometry detector (GC/MS). The water column particulate was also evaluated by ultrasonic extraction, followed by quantification by GC/MS. The results reveal that some of the compounds were found and it was due to anthropic activities in the vicinity of the regions. By initial analysis it was possible verify river that cross the RMB area present values below 0.05 μg L{sup -1}. (author)

  1. Measurement of reactivity worths of burnable poison rods in enriched uranium graphite-moderated core simulated to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Akino, Fujiyoshi; Takeuchi, Motoyoshi; Kitadate, Kenji; Yoshifuji, Hisashi; Kaneko, Yoshihiko

    1980-11-01

    As the core design for the Experimental Very High Temperature Gas Cooled Reactor progresses, evaluation of design precision has become increasingly important. For a high precision design, it is required to have adequate group constants based on accurate nuclear data, as well as calculation methods properly describing the physical behavior of neutrons. We, therefore, assembled a simulation core for VHTR, SHE-14, using a graphite-moderated 20%-enriched uranium Semi-Homogeneous Experimental Critical Facility (SHE), and obtained useful experimental data in evaluating the design precision. The VHTR is designed to accommodate burnable poison and control rods for reactivity compensation. Accordingly, the experimental burnable poison rods which are similar to those to be used in the experimental reactor were prepared, and their reactivity values were measured in the SHE-14 core. One to three rods of the above experimental burnable poison rods were inserted into the central column of the SHE-14 core, and the reactivity values were measured by the period and fuel rod substitution method. The results of the measurements have clearly shown that due to the self-shielding effect of B 4 C particles the reactivity value decreases with increasing particle diameter. For the particle diameter, the reactivity value is found to increase linearly with the logarithm of boron content. The measured values and those calculated are found to agree with each other within 5%. These results indicate that the reactivity of the burnable poison rod can be estimated fairly accurately by taking into account the self-shielding effect of B 4 C particles and the heterogeneity of the lattice cell. (author)

  2. Development and evaluation of a radial anaerobic/aerobic reactor treating organic matter and nitrogen in sewage

    Directory of Open Access Journals (Sweden)

    L. H. P. Garbossa

    2005-12-01

    Full Text Available The design and performance of a radial anaerobic/aerobic immobilized biomass (RAAIB reactor operating to remove organic matter, solids and nitrogen from sewage are discussed. The bench-scale RAAIB was divided into five concentric chambers. The second and fourth chambers were packed with polyurethane foam matrices. The performance of the reactor in removing organic matter and producing nitrified effluent was good, and its configuration favored the transfer of oxygen to the liquid mass due to its characteristics and the fixed polyurethane foam bed arrangement in concentric chambers. Partial denitrification of the liquid also took place in the RAAIB. The reactor achieved an organic matter removal efficiency of 84%, expressed as chemical oxygen demand (COD, and a total Kjeldahl nitrogen (TKN removal efficiency of 96%. Average COD, nitrite and nitrate values for the final effluent were 54 mg.L-1, 0.3 mg.L-1 and 22.1 mg.L-1, respectively.

  3. A feasibility study on long-life reduced-moderation water reactor with highly protected Pu breeding by doping with minor actinides

    International Nuclear Information System (INIS)

    Hamase, Erina; Damian, Frederic; Poinot-salanon, Christine; Saito, Masaki; Sagara, Hiroshi; Han, Chi Young

    2013-01-01

    Highlights: ► The present paper is focused on a reduced-moderation water reactor. ►237 Np or 241 Am was doped into the inner blanket. ►238 Pu transmuted from MA works as a fissionable nuclide. ► It also contributes to increasing the proliferation resistance of Pu. ► Long-life core and highly protected Pu breeding were simultaneously achieved. - Abstract: The present paper is focused on the feasibility of reduced-moderation water reactor (RMWR) which could simultaneously achieve the extension of core life-time, and highly protected plutonium (Pu) breeding performance with short compound system doubling time (CSDT) by neptunium-237 ( 237 Np) or americium-241 ( 241 Am) doping of the inner blanket of the RMWR. As a preliminary analysis of the RMWR, a simplified fuel pin configuration was analyzed. In case of 60% doping of 237 Np and 241 Am in the inner blanket, the maximum available effective full power days (EFPDs) were much longer and were about 15,100 and 14,200 EFPDs, respectively. For the proliferation resistance of Pu produced in the blanket, the total Pu in the inner and lower/upper blanket was below the practically unusable criterion for an explosive device proposed by Pellaud, and was technically unfeasible for high-technology hypothetical nuclear explosive devices (HNEDs) proposed by Kessler and Kimura. The protected Pu breeding performance was analyzed and in case of 60% doping of 237 Np and 241 Am, CSDT was short by approximately 40 and 50 years. Therefore, the feasibility of RMWR was shown which could simultaneously achieve the extension of core life-time and have highly protected Pu breeding performance with short CSDT by doping the inner blanket with minor actinides (MAs). The objective of this study was thus to evaluate the performance of the concept with respect to the core life-time, proliferation resistance of Pu and CSDT. The technological feasibility of such core concept will have to be evaluated by further dedicated analyses

  4. Socioeconomic status moderates age-related differences in the brain's functional network organization and anatomy across the adult lifespan.

    Science.gov (United States)

    Chan, Micaela Y; Na, Jinkyung; Agres, Phillip F; Savalia, Neil K; Park, Denise C; Wig, Gagan S

    2018-05-14

    An individual's environmental surroundings interact with the development and maturation of their brain. An important aspect of an individual's environment is his or her socioeconomic status (SES), which estimates access to material resources and social prestige. Previous characterizations of the relation between SES and the brain have primarily focused on earlier or later epochs of the lifespan (i.e., childhood, older age). We broaden this work to examine the relationship between SES and the brain across a wide range of human adulthood (20-89 years), including individuals from the less studied middle-age range. SES, defined by education attainment and occupational socioeconomic characteristics, moderates previously reported age-related differences in the brain's functional network organization and whole-brain cortical structure. Across middle age (35-64 years), lower SES is associated with reduced resting-state system segregation (a measure of effective functional network organization). A similar but less robust relationship exists between SES and age with respect to brain anatomy: Lower SES is associated with reduced cortical gray matter thickness in middle age. Conversely, younger and older adulthood do not exhibit consistent SES-related difference in the brain measures. The SES-brain relationships persist after controlling for measures of physical and mental health, cognitive ability, and participant demographics. Critically, an individual's childhood SES cannot account for the relationship between their current SES and functional network organization. These findings provide evidence that SES relates to the brain's functional network organization and anatomy across adult middle age, and that higher SES may be a protective factor against age-related brain decline. Copyright © 2018 the Author(s). Published by PNAS.

  5. Centralized Reliability Data Organization (CREDO) assessment of critical component unavailability in liquid metal reactors

    International Nuclear Information System (INIS)

    Koger, K.H.; Haire, M.J.; Humphrys, B.L.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.

    1988-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Dept. of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 20,000 components and approximately 1500 event records. A conservative estimation is that the total component operating hours is approaching 2.2 billion hours. The work reported here focuses on the availability information contained in CREDO and the development of availability critical items lists. That is, individual components are ranked in prioritized lists from worst to best performers from an availability standpoint. Availability as used here is an inherent characteristics of the component and is not necessarily related to plant operability. A major observation is that a few components have a much higher unavailability factor than the average. The top fifteen components contribute 93%, 77%, and 87% of the total system unavailability for EBR-II, FFTF, and JOYO respectively. Critical components common to all three sites are mechanical pumps and electromagnetic pumps. Application of resources to these components with the highest unavailability will have the greatest effect on overall availability. All three sites demonstrate that low maintainability (i.e., long repair times), rather than unreliability (i.e., high failure rates), are the main contributors, by about a two-to-one margin, to liquid metal system unavailability

  6. Organization of remote control of spectrometers at the IBR-2M reactor

    International Nuclear Information System (INIS)

    Kirilov, A.S.; Murashkevich, S.M.; Okulov, R.Yu.; Petukhova, T.B.

    2008-01-01

    The development of network infrastructure in Dubna creates conditions for the remote supervision/control of experiment at the IBR-2M reactor beyond the JINR local network. The opportunity of observing an experiment from outside makes it possible to respond to errors or unforeseen, to save the reactor time. The principles of organization and peculiarities of implementation of the WebSonix system consisting of the central web-site and communication facilities for spectrometers are considered in the paper. The system allows one to reflect the actual condition of all the components of the instrument, survey log files, visualize the accumulated spectra and to control the experimental procedure on the instruments controlled by the software complex Sonix+ (OS Windows XP). The system is an independent instrument, it is easy to extend or change and easy to adapt to the instrument data specifics. The system is implemented using the PHP and Python scripts. The OS GNU/Linux Debian and web-server Apache 2 are installed on the web-site computer

  7. Real-time measurements of secondary organic aerosol formation and aging from ambient air in an oxidation flow reactor in the Los Angeles area

    Directory of Open Access Journals (Sweden)

    A. M. Ortega

    2016-06-01

    Full Text Available Field studies in polluted areas over the last decade have observed large formation of secondary organic aerosol (SOA that is often poorly captured by models. The study of SOA formation using ambient data is often confounded by the effects of advection, vertical mixing, emissions, and variable degrees of photochemical aging. An oxidation flow reactor (OFR was deployed to study SOA formation in real-time during the California Research at the Nexus of Air Quality and Climate Change (CalNex campaign in Pasadena, CA, in 2010. A high-resolution aerosol mass spectrometer (AMS and a scanning mobility particle sizer (SMPS alternated sampling ambient and reactor-aged air. The reactor produced OH concentrations up to 4 orders of magnitude higher than in ambient air. OH radical concentration was continuously stepped, achieving equivalent atmospheric aging of 0.8 days–6.4 weeks in 3 min of processing every 2 h. Enhancement of organic aerosol (OA from aging showed a maximum net SOA production between 0.8–6 days of aging with net OA mass loss beyond 2 weeks. Reactor SOA mass peaked at night, in the absence of ambient photochemistry and correlated with trimethylbenzene concentrations. Reactor SOA formation was inversely correlated with ambient SOA and Ox, which along with the short-lived volatile organic compound correlation, indicates the importance of very reactive (τOH  ∼  0.3 day SOA precursors (most likely semivolatile and intermediate volatility species, S/IVOCs in the Greater Los Angeles Area. Evolution of the elemental composition in the reactor was similar to trends observed in the atmosphere (O : C vs. H : C slope  ∼  −0.65. Oxidation state of carbon (OSc in reactor SOA increased steeply with age and remained elevated (OSC  ∼  2 at the highest photochemical ages probed. The ratio of OA in the reactor output to excess CO (ΔCO, ambient CO above regional background vs. photochemical age is similar to

  8. Real-time measurements of secondary organic aerosol formation and aging from ambient air in an oxidation flow reactor in the Los Angeles area

    Science.gov (United States)

    Ortega, Amber M.; Hayes, Patrick L.; Peng, Zhe; Palm, Brett B.; Hu, Weiwei; Day, Douglas A.; Li, Rui; Cubison, Michael J.; Brune, William H.; Graus, Martin; Warneke, Carsten; Gilman, Jessica B.; Kuster, William C.; de Gouw, Joost; Gutiérrez-Montes, Cándido; Jimenez, Jose L.

    2016-06-01

    Field studies in polluted areas over the last decade have observed large formation of secondary organic aerosol (SOA) that is often poorly captured by models. The study of SOA formation using ambient data is often confounded by the effects of advection, vertical mixing, emissions, and variable degrees of photochemical aging. An oxidation flow reactor (OFR) was deployed to study SOA formation in real-time during the California Research at the Nexus of Air Quality and Climate Change (CalNex) campaign in Pasadena, CA, in 2010. A high-resolution aerosol mass spectrometer (AMS) and a scanning mobility particle sizer (SMPS) alternated sampling ambient and reactor-aged air. The reactor produced OH concentrations up to 4 orders of magnitude higher than in ambient air. OH radical concentration was continuously stepped, achieving equivalent atmospheric aging of 0.8 days-6.4 weeks in 3 min of processing every 2 h. Enhancement of organic aerosol (OA) from aging showed a maximum net SOA production between 0.8-6 days of aging with net OA mass loss beyond 2 weeks. Reactor SOA mass peaked at night, in the absence of ambient photochemistry and correlated with trimethylbenzene concentrations. Reactor SOA formation was inversely correlated with ambient SOA and Ox, which along with the short-lived volatile organic compound correlation, indicates the importance of very reactive (τOH ˜ 0.3 day) SOA precursors (most likely semivolatile and intermediate volatility species, S/IVOCs) in the Greater Los Angeles Area. Evolution of the elemental composition in the reactor was similar to trends observed in the atmosphere (O : C vs. H : C slope ˜ -0.65). Oxidation state of carbon (OSc) in reactor SOA increased steeply with age and remained elevated (OSC ˜ 2) at the highest photochemical ages probed. The ratio of OA in the reactor output to excess CO (ΔCO, ambient CO above regional background) vs. photochemical age is similar to previous studies at low to moderate ages and also extends to

  9. Single-reactor process for producing liquid-phase organic compounds from biomass

    Science.gov (United States)

    Dumesic, James A [Verona, WI; Simonetti, Dante A [Middleton, WI; Kunkes, Edward L [Madison, WI

    2011-12-13

    Disclosed is a method for preparing liquid fuel and chemical intermediates from biomass-derived oxygenated hydrocarbons. The method includes the steps of reacting in a single reactor an aqueous solution of a biomass-derived, water-soluble oxygenated hydrocarbon reactant, in the presence of a catalyst comprising a metal selected from the group consisting of Cr, Mn, Fe, Co, Ni, Cu, Mo, Tc, Ru, Rh, Pd, Ag, W, Re, Os, Ir, Pt, and Au, at a temperature, and a pressure, and for a time sufficient to yield a self-separating, three-phase product stream comprising a vapor phase, an organic phase containing linear and/or cyclic mono-oxygenated hydrocarbons, and an aqueous phase.

  10. Influence of the type of organisms on the biomass hold-up in a fluidized-bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Timmermans, P.; Haute, A. van

    1984-01-01

    In the last few years, the use of fluidized-bed reactors for biological wastewater treatment has got increasing attention. In 1981, Shieh et al. proposed a model to predict the biomass concentration in a fluidized-bed reactor. From this model one can see that the biofilm density plays a very important role in determining the total biomass hold-up. In this article the influence of the type of carbon source on the biomass concentration, and as a consequence the type of organisms selected, is studied. The growth of a filamentous, budforming bacteria in a reactor treating nitrate rich surface water supplied with methanol as carbon source, results in a biomass concentration only half of the concentration which can normally be obtained in a fluidized-bed reactor treating synthetic wastewater; in this latter case rod-shaped bacteria are enriched which permit a dense packing.

  11. Oxidation of organics in water in microfluidic electrochemical reactors: Theoretical model and experiments

    International Nuclear Information System (INIS)

    Scialdone, Onofrio; Guarisco, Chiara; Galia, Alessandro

    2011-01-01

    The electrochemical oxidation of organics in water performed in micro reactors on boron doped diamond (BDD) anode was investigated both theoretically and experimentally in order to find the influence of various operative parameters on the conversion and the current efficiency CE of the process. The electrochemical oxidation of formic acid (FA) was selected as a model case. High conversions for a single passage of the electrolytic solution inside the cell were obtained by operating with proper residence times and low distances between cathode and anode. The effect of initial concentration, flow rate and current density was investigated in detail. Theoretical predictions were in very good agreement with experimental results for both mass transfer control, oxidation reaction control and mixed kinetic regimes in spite of the fact that no adjustable parameters was used. Mass transfer process was successfully modelled by considering for simplicity a constant Sh number (e.g., a constant mass transfer coefficient k m ) for a process performed with no high values of the current intensity to minimize the effect of the gas bubbling on the flowdynamic pattern. For mixed kinetic regimes, two different modelling approaches were used. In the first one, the oxidation of organics at BDD was assumed to be mass transfer controlled and to occur with an intrinsic 100% CE when applied current density is higher than the limiting current density. In the second case, the CE of the process was modelled assuming that the competition between organic and water oxidation depends only on the electrodic material and on the nature and the concentration of the organic. In the latter case a better agreement between experimental data and theoretical predictions was observed.

  12. Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

    KAUST Repository

    Zhu, Xiuping; Logan, Bruce E.

    2013-01-01

    accomplish H2O2 generation and Fe2+ release for the Fenton reaction. In tests using phenol, 75±2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22h), and phenol was completely degraded to simple and readily

  13. Industrial complex in organizing the high-speed in-line construction of reactor compartments at the Balakovo NPP

    International Nuclear Information System (INIS)

    Maksakov, A.I.; Kovrigin, Yu.K.; Zhila, V.P.

    1986-01-01

    Qualitatively new technology of reactor compartment construction presupposing organizing of an industrial-mounting in-line complex is described. Maximum level of construction industrialization and noticeable reduction of construction duration are noted to be ensured by means of this technology

  14. In situ secondary organic aerosol formation from ambient pine forest air using an oxidation flow reactor

    Science.gov (United States)

    Palm, Brett B.; Campuzano-Jost, Pedro; Ortega, Amber M.; Day, Douglas A.; Kaser, Lisa; Jud, Werner; Karl, Thomas; Hansel, Armin; Hunter, James F.; Cross, Eben S.; Kroll, Jesse H.; Peng, Zhe; Brune, William H.; Jimenez, Jose L.

    2016-03-01

    An oxidation flow reactor (OFR) is a vessel inside which the concentration of a chosen oxidant can be increased for the purpose of studying SOA formation and aging by that oxidant. During the BEACHON-RoMBAS (Bio-hydro-atmosphere interactions of Energy, Aerosols, Carbon, H2O, Organics & Nitrogen-Rocky Mountain Biogenic Aerosol Study) field campaign, ambient pine forest air was oxidized by OH radicals in an OFR to measure the amount of SOA that could be formed from the real mix of ambient SOA precursor gases, and how that amount changed with time as precursors changed. High OH concentrations and short residence times allowed for semicontinuous cycling through a large range of OH exposures ranging from hours to weeks of equivalent (eq.) atmospheric aging. A simple model is derived and used to account for the relative timescales of condensation of low-volatility organic compounds (LVOCs) onto particles; condensational loss to the walls; and further reaction to produce volatile, non-condensing fragmentation products. More SOA production was observed in the OFR at nighttime (average 3 µg m-3 when LVOC fate corrected) compared to daytime (average 0.9 µg m-3 when LVOC fate corrected), with maximum formation observed at 0.4-1.5 eq. days of photochemical aging. SOA formation followed a similar diurnal pattern to monoterpenes, sesquiterpenes, and toluene+p-cymene concentrations, including a substantial increase just after sunrise at 07:00 local time. Higher photochemical aging (> 10 eq. days) led to a decrease in new SOA formation and a loss of preexisting OA due to heterogeneous oxidation followed by fragmentation and volatilization. When comparing two different commonly used methods of OH production in OFRs (OFR185 and OFR254-70), similar amounts of SOA formation were observed. We recommend the OFR185 mode for future forest studies. Concurrent gas-phase measurements of air after OH oxidation illustrate the decay of primary VOCs, production of small oxidized organic

  15. In situ secondary organic aerosol formation from ambient pine forest air using an oxidation flow reactor

    Directory of Open Access Journals (Sweden)

    B. B. Palm

    2016-03-01

    Full Text Available An oxidation flow reactor (OFR is a vessel inside which the concentration of a chosen oxidant can be increased for the purpose of studying SOA formation and aging by that oxidant. During the BEACHON-RoMBAS (Bio-hydro-atmosphere interactions of Energy, Aerosols, Carbon, H2O, Organics & Nitrogen–Rocky Mountain Biogenic Aerosol Study field campaign, ambient pine forest air was oxidized by OH radicals in an OFR to measure the amount of SOA that could be formed from the real mix of ambient SOA precursor gases, and how that amount changed with time as precursors changed. High OH concentrations and short residence times allowed for semicontinuous cycling through a large range of OH exposures ranging from hours to weeks of equivalent (eq. atmospheric aging. A simple model is derived and used to account for the relative timescales of condensation of low-volatility organic compounds (LVOCs onto particles; condensational loss to the walls; and further reaction to produce volatile, non-condensing fragmentation products. More SOA production was observed in the OFR at nighttime (average 3 µg m−3 when LVOC fate corrected compared to daytime (average 0.9 µg m−3 when LVOC fate corrected, with maximum formation observed at 0.4–1.5 eq. days of photochemical aging. SOA formation followed a similar diurnal pattern to monoterpenes, sesquiterpenes, and toluene+p-cymene concentrations, including a substantial increase just after sunrise at 07:00 local time. Higher photochemical aging (> 10 eq. days led to a decrease in new SOA formation and a loss of preexisting OA due to heterogeneous oxidation followed by fragmentation and volatilization. When comparing two different commonly used methods of OH production in OFRs (OFR185 and OFR254-70, similar amounts of SOA formation were observed. We recommend the OFR185 mode for future forest studies. Concurrent gas-phase measurements of air after OH oxidation illustrate the decay of primary VOCs, production

  16. Removal of Organic Load in Communal Wastewater by using the Six Stage Anaerobic Baffle Reactor (ABR

    Directory of Open Access Journals (Sweden)

    Trilita Minarni Nur

    2016-01-01

    Full Text Available The reduction of water quality in the urban drainage is a crucial problem to overcome because it can affect the health of community. This fact encouraged the researcher to conduct the research in efforts to increase the water quality in the drainage. One of the solutions to increase the water quality in the drainage is that the domestic wastewater must be treated at first before it is flown through the drainage. Furthermore, the wastewater treatment was conducted by employing the communal wastewater processor. The research was aimed at knowing the capability of Anaerobic Baffle Reactor with the six-stage design in communal wastewater processor in efforts to decrease the organic load. This research was conducted in a laboratory scale. Meanwhile, the sort of waste used was taken from the domestic wastewater of settlement by varying its discharge and waste concentration flowing into the waste processor. Finally, the research result showed that the reduction of organic load of COD was reaching up to 92%, N was 85% and Phosphate was 50%.

  17. STUDY OF WATER HAMMERS IN THE FILLING OF THE SYSTEM OF PRESSURE COMPENSATION IN THE WATER-COOLED AND WATER-MODERATED POWER REACTORS

    Directory of Open Access Journals (Sweden)

    A. V. Korolyev

    2017-01-01

    list of initial events of severe accidents at NPPs with a water-cooled and water-moderated power reactor can be expanded.

  18. Organic loading rate effect on the acidogenesis of cheese whey: a comparison between UASB and SBR reactors.

    Science.gov (United States)

    Calero, R; Iglesias-Iglesias, R; Kennes, C; Veiga, M C

    2017-09-16

    Volatile fatty acids (VFA) production and degree of acidification (DA) were investigated in the anaerobic treatment of cheese whey by comparison of two processes: a continuous process using a laboratory upflow anaerobic sludge blanket (UASB) reactor and a discontinuous process using a sequencing batch reactor (SBR). The main purpose of this work was to study the organic loading rate (OLR) effect on the yield of VFA in two kinds of reactors. The predominant products in the acidogenic process in both reactors were: acetate, propionate, butyrate and valerate. The maximum DA obtained was 98% in an SBR at OLR of 2.7 g COD L -1 d -1 , and 97% in the UASB at OLR at 15.1 g COD L -1 d -1 . The results revealed that the UASB reactor was more efficient at a medium OLR with a higher VFA yield, while with the SBR reactor, the maximum acidification was obtained at a lower OLR with changes in the VFA profile at different OLRs applied.

  19. Low concentration volatile organic pollutants removal in combined adsorber-desorber-catalytic reactor system

    Directory of Open Access Journals (Sweden)

    Arsenijević Zorana

    2008-01-01

    Full Text Available The removal of volatile organic compounds (VOCs from numerous emission sources is of crucial importance due to more rigorous demands on air quality. Different technologies can be used to treat the VOCs from effluent gases: absorption, physical adsorption, open flame combustion, thermal and catalytic incineration. Their appropriateness for the specific process depends on several factors such as efficiency, energy consumption, secondary pollution, capital investments etc. The distinctive features of the catalytic combustion are high efficiency and selectivity toward be­nign products, low energy consumption and absence of secondary polluti­on. The supported noble catalysts are widely used for catalytic incineration due to their low ignition temperatures and high thermal and chemical stability. In our combined system adsorption and desorption are applied in the spouted bed with draft tube (SBDT unit. The annular zone, loaded with sorbent, was divided in adsorption and desorption section. Draft tube enabled sorbent recirculation between sections. Combustion of desorbed gases to CO2 and water vapor are realized in additive catalytic reactor. This integrated device provided low concentrations VOCs removal with reduced energy consumption. Experiments were conducted on a pilot unit of 220 m3/h nominal capacity. The sorbent was activated carbon, type K81/B - Trayal Corporation, Krusevac. A sphere shaped commercial Pt/Al2O3 catalyst with "egg-shell" macro-distribution was used for the investigation of xylene deep oxidation. Within this paper the investigations of removal of xylene vapors, a typical pollutant in production of liquid pesticides, in combined adsorber/desorber/catalytic reactor system is presented.

  20. Efficiency of a pilot-scale integrated sludge thickening and digestion reactor in treating low-organic excess sludge.

    Science.gov (United States)

    He, Qiang; Li, Jiang; Liu, Hongxia; Tang, Chuandong; de Koning, Jaap; Spanjers, Henri

    2012-06-01

    The sludge production from medium- and small-scale wastewater treatment plants in the Three Gorges Reservoir Region is low and non-stable; especially, the organic content in this sludge is low (near 40% of VS/TS). An integrated thickening and digestion (ISTD) reactor was developed to treat this low-organic excess sludge. After a flow test and start-up experiment of the reactor, a running experiment was used to investigate the excess sludge treatment efficiency under five different excess sludge inflows: 200, 300, 400, 500 and 400 L/d (a mixture of excess sludge and primary sludge in a volume ratio of 9:1). This trial was carried out in the wastewater treatment plant in Chongqing, which covers 80% of the Three Gorges Reservoir Region, under the following conditions: (1) sludge was heated to 38-40 degrees C using an electrical heater to maintain anaerobic mesophilic digestion; (2) the biogas produced was recirculated to mix raw sludge with anaerobic sludge in the reactor under the flow rate of 12.5 L/min. There were three main results. Firstly, the flow pattern of the inner reactor was almost completely mixed under the air flow of 12.0 L/min using clear water. Secondly, under all the different sludge inflows, the water content in the outlet sludge was below 93%. Thirdly, the organic content in the outlet sludge was decreased from 37% to 30% and from 24% to 20%, whose removal ratio was in relation to the organic content of the inlet sludge. The excess sludge treatment capacity of the ISTD reactor was according to the organic content in the excess sludge.

  1. 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

  2. Organics removal from landfill leachate and activated sludge production in SBR reactors

    International Nuclear Information System (INIS)

    Klimiuk, Ewa; Kulikowska, Dorota

    2006-01-01

    This study is aimed at estimating organic compounds removal and sludge production in SBR during treatment of landfill leachate. Four series were performed. At each series, experiments were carried out at the hydraulic retention time (HRT) of 12, 6, 3 and 2 d. The series varied in SBR filling strategies, duration of the mixing and aeration phases, and the sludge age. In series 1 and 2 (a short filling period, mixing and aeration phases in the operating cycle), the relationship between organics concentration (COD) in the leachate treated and HRT was pseudo-first-order kinetics. In series 3 (with mixing and aeration phases) and series 4 (only aeration phase) with leachate supplied by means of a peristaltic pump for 4 h of the cycle (filling during reaction period) - this relationship was zero-order kinetics. Activated sludge production expressed as the observed coefficient of biomass production (Y obs ) decreased correspondingly with increasing HRT. The smallest differences between reactors were observed in series 3 in which Y obs was almost stable (0.55-0.6 mg VSS/mg COD). The elimination of the mixing phase in the cycle (series 4) caused the Y obs to decrease significantly from 0.32 mg VSS/mg COD at HRT 2 d to 0.04 mg VSS/mg COD at HRT 12 d. The theoretical yield coefficient Y accounted for 0.534 mg VSS/mg COD (series 1) and 0.583 mg VSS/mg COD (series 2). In series 3 and 4, it was almost stable (0.628 mg VSS/mg COD and 0.616 mg VSS/mg COD, respectively). After the elimination of the mixing phase in the operating cycle, the specific biomass decay rate increased from 0.006 d -1 (series 3) to 0.032 d -1 (series 4). The operating conditions employing mixing/aeration or only aeration phases enable regulation of the sludge production. The SBRs operated under aerobic conditions are more favourable at a short hydraulic retention time. At long hydraulic retention time, it can lead to a decrease in biomass concentration in the SBR as a result of cell decay. On the contrary

  3. Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, light water moderator and beryllium reflectors

    International Nuclear Information System (INIS)

    Liem, Peng Hong; Sembiring, Tagor Malem

    2012-01-01

    Highlights: ► Benchmark calculations of the new JENDL-4.0 library. ► Thermal research reactor with oxide LEU fuel, H 2 O moderator and Be reflector. ► JENDL-4.0 library shows better C/E values for criticality evaluations. - Abstract: Benchmark calculations of the new JENDL-4.0 library on the criticality experiments of a thermal research reactor with oxide low enriched uranium (LEU, 20 w/o) fuel, light water moderator and beryllium reflector (RSG GAS) have been conducted using a continuous energy Monte Carlo code, MVP-II. The JENDL-4.0 library shows better C/E values compared to the former library JENDL-3.3 and other world-widely used latest libraries (ENDF/B-VII.0 and JEFF-3.1).

  4. Louisiana SIP: LAC 33:III Ch 21 Subchap J, 2147--Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 1998-02-02 (LAc74) to more..

    Science.gov (United States)

    Louisiana SIP: LAC 33:III Ch 21 Subchap J, 2147--Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 1998-02-02 (LAc74) more...

  5. Louisiana SIP: LAC 33:III Ch 2147. Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 2011-08-04 (LAd34) to 2017-09-27

    Science.gov (United States)

    Louisiana SIP: LAC 33:III Ch 2147. Limiting Volatile Organic Compound (VOC) Emissions from Reactor Processes and Distillation Operations in Synthetic Organic Chemical manufacturing Industry (SOCMI); SIP effective 2011-08-04 (LAd34) to 2017-09-27

  6. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  7. Patterns identification in supervisory systems of nuclear reactors installations and gas pipelines systems using self-organizing maps

    International Nuclear Information System (INIS)

    Doraskevicius Junior, Waldemar

    2005-01-01

    Self-Organizing Maps, SOM, of Kohonen were studied, implemented and tested with the aim of developing, for the energy branch, an effective tool especially for transient identification in nuclear reactors and for gas pipelines networks logistic supervision, by classifying operations and identifying transients or abnormalities. The digital system for the test was developed in Java platform, for the portability and scalability, and for belonging to free development platforms. The system, executed in personal computers, showed satisfactory results to aid in decision taking, by classifying IRIS (International Reactor Innovative and Secure) reactor operation conditions (data from simulator) and by classifying Southeast (owner: TRANSPETRO - Brazil) gas pipeline network. Various adaptations were needed for such business, as new topologies for the output layer of artificial neural network and particular preparation for the input data. (author)

  8. Organization and management of health physics support for a research reactor

    International Nuclear Information System (INIS)

    Bates, E.F.; Neff, R.D.; Randall, J.D.

    1980-01-01

    The Radiological Safety Office administers the radiological safety and surveillance programs for Texas A and M University. This program includes the assignment of a health physics group to the Texas A and M University Nuclear Science Center. By mutual agreement, the Nuclear Science Center health physics group acts as an integral part of the NSC staff which provides a system for making positive contributions to the decision-making process and the management of its time and resources to accomplish the design objectives of the radiation safety program. These personnel administer a continuous program of hazard analyses and evaluations to minimize and eliminate radiological hazards in terms of occupational exposures to radiation, contamination control, and release of radioactive effluents to the environs. This program has been effective in reducing occupational exposures to radiation in terms of total manrem expended and maintaining effluent releases to the environment at approximately 2% of the limits specified in 10CFR20. This paper presents' an organizational method for establishing an operational and functional research reactor health physics group and the resultant benefits from its contribution to the overall organization. (author)

  9. The European reliability data system. An organized information exchange on the operation of European nuclear reactors

    International Nuclear Information System (INIS)

    Mancini, G.; Amesz, J.; Bastianini, P.; Capobianchi, S.

    1983-01-01

    The paper revises the aims and objectives of the European Reliability Data System (ERDS), a centralized system collecting and organizing, at European level, information related to the operation of LWRs. The ERDS project was started in 1977 and after a preliminary feasibility study that ended in 1979 is now proceeding towards the final design and implementation stages. ERDS exploits information collected in national data systems and information deriving from single reactor sources. The paper describes first the development of the four data banks constituting the system: Component Event Data Bank, CEDB; Abnormal Occurrences Reporting System, AORS; Operating Unit Status Report, OUSR; and Generic Reliability Parameter Data Bank, GRPDB. Several typical aspects concerning the project are then outlined from the need of homogeneization of data and therefore the need for setting up reference classifications, to the problem of data transcoding and input into the system. Furthermore, the need is stressed of involving much more deeply nuclear power plant operators into the process of data acquisition by providing them with a useful feedback from the data analysis. (author)

  10. Artificial neural network with self-organizing mapping for reactor stability monitoring

    International Nuclear Information System (INIS)

    Okumura, Motofumi; Tsuji, Masashi; Shimazu, Yoichiro

    2009-01-01

    In boiling water reactor (BWR) stability monitoring, damping ratio has been used as a stability index. A method for estimating the damping ratio by applying Principal Component Analysis (PCA) to neutron detector signals measured with local power range monitors (LPRMs) had been developed; in this method, measured fluctuating signal is decomposed into some independent components and the signal components directly related to stability are extracted among them to determine the damping ratio. For online monitoring, it is necessary to select stability related signal components efficiently. The self-organizing map (SOM) is one of the artificial neural networks (ANNs) and has the characteristics such that online learning is possible without supervised learning within a relatively short time. In the present study, the SOM was applied to extract the relevant signal components more quickly and more accurately, and the availability was confirmed through the feasibility study. For realizing online stability monitoring only with ANNs, another type of ANN that performs online processing of PCA was combined with SOM. And stability monitoring performance was investigated. (author)

  11. Distributions and activities of ammonia oxidizing bacteria and polyphosphate accumulating organisms in a pumped-flow biofilm reactor.

    Science.gov (United States)

    Wu, Guangxue; Nielsen, Michael; Sorensen, Ketil; Zhan, Xinmin; Rodgers, Michael

    2009-10-01

    The spatial distributions and activities of ammonia oxidizing bacteria (AOB) and polyphosphate accumulating organisms (PAOs) were investigated for a novel laboratory-scale sequencing batch pumped-flow biofilm reactor (PFBR) system that was operated for carbon, nitrogen and phosphorus removal. The PFBR comprised of two 16.5l tanks (Reactors 1 and 2), each with a biofilm module of 2m(2) surface area. To facilitate the growth of AOB and PAOs in the reactor biofilms, the influent wastewater was held in Reactor 1 under stagnant un-aerated conditions for 6 h after feeding, and was then pumped over and back between Reactors 1 and 2 for 12 h, creating aerobic conditions in the two reactors during this period; as a consequence, the biofilm in Reactor 2 was in an aerobic environment for almost all the 18.2 h operating cycle. A combination of micro-sensor measurements, molecular techniques, batch experiments and reactor studies were carried out to analyse the performance of the PFBR system. After 100 days operation at a filtered chemical oxygen demand (COD(f)) loading rate of 3.46 g/m(2) per day, the removal efficiencies were 95% COD(f), 87% TN(f) and 74% TP(f). While the PFBR microbial community structure and function were found to be highly diversified with substantial AOB and PAO populations, about 70% of the phosphorus release potential and almost 100% of the nitrification potential were located in Reactors 1 and 2, respectively. Co-enrichment of AOB and PAOs was realized in the Reactor 2 biofilm, where molecular analyses revealed unexpected microbial distributions at micro-scale, with population peaks of AOB in a 100-250 microm deep sub-surface zone and of PAOs in the 0-150 microm surface zone. The micro-distribution of AOB coincided with the position of the nitrification peak identified during micro-sensor analyses. The study demonstrates that enrichment of PAOs can be realized in a constant or near constant aerobic biofilm environment. Furthermore, the findings suggest

  12. Secondary organic aerosol formation from in-use motor vehicle emissions using a potential aerosol mass reactor.

    Science.gov (United States)

    Tkacik, Daniel S; Lambe, Andrew T; Jathar, Shantanu; Li, Xiang; Presto, Albert A; Zhao, Yunliang; Blake, Donald; Meinardi, Simone; Jayne, John T; Croteau, Philip L; Robinson, Allen L

    2014-10-07

    Secondary organic aerosol (SOA) formation from in-use vehicle emissions was investigated using a potential aerosol mass (PAM) flow reactor deployed in a highway tunnel in Pittsburgh, Pennsylvania. Experiments consisted of passing exhaust-dominated tunnel air through a PAM reactor over integrated hydroxyl radical (OH) exposures ranging from ∼ 0.3 to 9.3 days of equivalent atmospheric oxidation. Experiments were performed during heavy traffic periods when the fleet was at least 80% light-duty gasoline vehicles on a fuel-consumption basis. The peak SOA production occurred after 2-3 days of equivalent atmospheric oxidation. Additional OH exposure decreased the SOA production presumably due to a shift from functionalization to fragmentation dominated reaction mechanisms. Photo-oxidation also produced substantial ammonium nitrate, often exceeding the mass of SOA. Analysis with an SOA model highlight that unspeciated organics (i.e., unresolved complex mixture) are a very important class of precursors and that multigenerational processing of both gases and particles is important at longer time scales. The chemical evolution of the organic aerosol inside the PAM reactor appears to be similar to that observed in the atmosphere. The mass spectrum of the unoxidized primary organic aerosol closely resembles ambient hydrocarbon-like organic aerosol (HOA). After aging the exhaust equivalent to a few hours of atmospheric oxidation, the organic aerosol most closely resembles semivolatile oxygenated organic aerosol (SV-OOA) and then low-volatility organic aerosol (LV-OOA) at higher OH exposures. Scaling the data suggests that mobile sources contribute ∼ 2.9 ± 1.6 Tg SOA yr(-1) in the United States, which is a factor of 6 greater than all mobile source particulate matter emissions reported by the National Emissions Inventory. This highlights the important contribution of SOA formation from vehicle exhaust to ambient particulate matter concentrations in urban areas.

  13. The Centralized Reliability Data Organization (CREDO); an advanced nuclear reactor reliability, availability, and maintainability data bank and data analysis center

    International Nuclear Information System (INIS)

    Knee, H.E.

    1991-01-01

    The Centralized Reliability Data Organization (CREDO) is a data bank and data analysis center, which since 1985 has been jointly sponsored by the US Department of Energy's (US DOE's) Office of Technology Support Programs and Japan's Power Reactor and Nuclear Fuel Development Corporation (PNC). It focuses on reliability, availability and maintainability (RAM) data for components (e.g. valves, pumps, etc.) operating in advanced nuclear reactor facilities. As originally intended, the purpose of the CREDO system was to provide a centralized source of accurate, up-to-date data and information for use in RAM analyses necessary for meeting DOE's data needs in the areas of advanced reactor safety assessments, design and licensing. In particular, creation of the CREDO system was considered an essential element needed to fulfill the DOE Breeder Reactor Safety Program's commitment of 'identifying and exploiting areas in which probabilistic methods can be developed and used in making reactor safety Research and Development choices and optimizing designs of safety systems'. CREDO and its operation are explained. (author)

  14. 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

  15. Implementation of a management system for operating organizations of research reactors

    International Nuclear Information System (INIS)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes

    2011-01-01

    This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

  16. Implementation of a management system for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo, E-mail: kibrit@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Aquino, Afonso Rodrigues de; Zouain, Desiree Moraes, E-mail: araquino@ipen.b, E-mail: dmzouain@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    This paper presents the requirements established by an IAEA draft technical document for the implementation of a management system for operating organisations of research reactors. The following aspects will be discussed: structure of IAEA draft technical document, management system requirements, processes common to all research reactors, aspects for the implementation of the management system, and a formula for grading the management system requirements. (author)

  17. Secondary organic aerosol formation from ambient air in an oxidation flow reactor in central Amazonia

    Science.gov (United States)

    Palm, Brett B.; de Sá, Suzane S.; Day, Douglas A.; Campuzano-Jost, Pedro; Hu, Weiwei; Seco, Roger; Sjostedt, Steven J.; Park, Jeong-Hoo; Guenther, Alex B.; Kim, Saewung; Brito, Joel; Wurm, Florian; Artaxo, Paulo; Thalman, Ryan; Wang, Jian; Yee, Lindsay D.; Wernis, Rebecca; Isaacman-VanWertz, Gabriel; Goldstein, Allen H.; Liu, Yingjun; Springston, Stephen R.; Souza, Rodrigo; Newburn, Matt K.; Lizabeth Alexander, M.; Martin, Scot T.; Jimenez, Jose L.

    2018-01-01

    Secondary organic aerosol (SOA) formation from ambient air was studied using an oxidation flow reactor (OFR) coupled to an aerosol mass spectrometer (AMS) during both the wet and dry seasons at the Observations and Modeling of the Green Ocean Amazon (GoAmazon2014/5) field campaign. Measurements were made at two sites downwind of the city of Manaus, Brazil. Ambient air was oxidized in the OFR using variable concentrations of either OH or O3, over ranges from hours to days (O3) or weeks (OH) of equivalent atmospheric aging. The amount of SOA formed in the OFR ranged from 0 to as much as 10 µg m-3, depending on the amount of SOA precursor gases in ambient air. Typically, more SOA was formed during nighttime than daytime, and more from OH than from O3 oxidation. SOA yields of individual organic precursors under OFR conditions were measured by standard addition into ambient air and were confirmed to be consistent with published environmental chamber-derived SOA yields. Positive matrix factorization of organic aerosol (OA) after OH oxidation showed formation of typical oxidized OA factors and a loss of primary OA factors as OH aging increased. After OH oxidation in the OFR, the hygroscopicity of the OA increased with increasing elemental O : C up to O : C ˜ 1.0, and then decreased as O : C increased further. Possible reasons for this decrease are discussed. The measured SOA formation was compared to the amount predicted from the concentrations of measured ambient SOA precursors and their SOA yields. While measured ambient precursors were sufficient to explain the amount of SOA formed from O3, they could only explain 10-50 % of the SOA formed from OH. This is consistent with previous OFR studies, which showed that typically unmeasured semivolatile and intermediate volatility gases (that tend to lack C = C bonds) are present in ambient air and can explain such additional SOA formation. To investigate the sources of the unmeasured SOA-forming gases during this campaign

  18. Secondary organic aerosol formation from ambient air in an oxidation flow reactor in central Amazonia

    Directory of Open Access Journals (Sweden)

    B. B. Palm

    2018-01-01

    Full Text Available Secondary organic aerosol (SOA formation from ambient air was studied using an oxidation flow reactor (OFR coupled to an aerosol mass spectrometer (AMS during both the wet and dry seasons at the Observations and Modeling of the Green Ocean Amazon (GoAmazon2014/5 field campaign. Measurements were made at two sites downwind of the city of Manaus, Brazil. Ambient air was oxidized in the OFR using variable concentrations of either OH or O3, over ranges from hours to days (O3 or weeks (OH of equivalent atmospheric aging. The amount of SOA formed in the OFR ranged from 0 to as much as 10 µg m−3, depending on the amount of SOA precursor gases in ambient air. Typically, more SOA was formed during nighttime than daytime, and more from OH than from O3 oxidation. SOA yields of individual organic precursors under OFR conditions were measured by standard addition into ambient air and were confirmed to be consistent with published environmental chamber-derived SOA yields. Positive matrix factorization of organic aerosol (OA after OH oxidation showed formation of typical oxidized OA factors and a loss of primary OA factors as OH aging increased. After OH oxidation in the OFR, the hygroscopicity of the OA increased with increasing elemental O : C up to O : C ∼ 1.0, and then decreased as O : C increased further. Possible reasons for this decrease are discussed. The measured SOA formation was compared to the amount predicted from the concentrations of measured ambient SOA precursors and their SOA yields. While measured ambient precursors were sufficient to explain the amount of SOA formed from O3, they could only explain 10–50 % of the SOA formed from OH. This is consistent with previous OFR studies, which showed that typically unmeasured semivolatile and intermediate volatility gases (that tend to lack C = C bonds are present in ambient air and can explain such additional SOA formation. To investigate the sources of the

  19. Evaluation of filters in RSPCS (Reactor Service Pool Cooling System) and HWL (Hot Water Layer) in OPAL research reactor at ANSTO (Australian Nuclear Science and Technology Organization) using Gamma Spectrometry System and Liquid Scintillation Counter

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jim In; Foy, Robin; Jung, Seong Moon; Park, Hyeon Suk; Ye, Sung Joon [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Australian Nuclear Science and Technology Organization(ANSTO) has a research reactor, OPAL (Open Pool Australian Lightwater reactor) which is a state-of-art 20 MW reactor for various purposes. In OPAL reactor, there are many kinds of radionuclides produced from various reactions in pool water and those should be identified and quantified for the safe use of OPAL. To do that, it is essential to check the efficiency of filters which are able to remove the radioactive substance from the reactor pool water. There are two main water circuits in OPAL which are RSPCS (Reactor Service Pool Cooling System) and HWL (Hot Water Layer) water circuits. The reactor service pool is connected to the reactor pool via a transfer canal and provides a working area and storage space for the spent and other materials. Also, HWL is the upper part of the reactor pool water and it minimize radiation dose rates at the pool surface. We collected water samples from these circuits and measured the radioactivity by using Gamma Spectrometry System (GSS) and Liquid Scintillation Counter (LSC) to evaluate the filters. We could evaluate the efficiency of filters in RSPCS and HWL in OPAL research reactor. Through the measurements of radioactivity using GSS and LSC, we could conclude that there is likely to be no alpha emitter in water samples, and for beta and gamma activity, there are very big differences between inlet and outlet results, so every filter is working efficiently to remove the radioactive substance.

  20. 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)

  1. Removal of Legacy Low-Level Waste Reactor Moderator De-ionizer Resins Highly Contaminated with Carbon-14 from the 'Waste with no Path to Disposal List' Through Innovative Technical Analysis and Performance Assessment Techniques

    International Nuclear Information System (INIS)

    Goldston, W.T.; Hiergesell, R.A.; Kaplan, D.I.; Pope, H.L.

    2006-01-01

    At the Savannah River Site (SRS), nuclear production reactors used de-ionizers to control the chemistry of the reactor moderator during their operation to produce nuclear materials primarily for the weapons program. These de-ionizers were removed from the reactors and stored as a legacy waste and due to the relatively high carbon-14 (C-14) contamination (i.e., on the order of 740 giga becquerel (GBq) (20 curies) per de-ionizer) were considered a legacy 'waste with no path to disposal'. Considerable progress has been made in consideration of a disposal path for the legacy reactor de-ionizers. Presently, 48 - 50 de-ionizers being stored at SRS have 'no path to disposal' because the disposal limit for C-14 in the SRS's low-level waste disposal facility's Intermediate Level Vault (ILV) is only 160 GBq (4.2 curies) per vault. The current C-14 ILV disposal limit is based on a very conservative analysis of the air pathway. The paper will describe the alternatives that were investigated that resulted in the selection of a route to pursue. This paper will then describe SRS's efforts to reduce the conservatism in the analysis, which resulted in a significantly larger C-14 disposal limit. The work consisted of refining the gas-phase analysis to simulate the migration of C-14 from the waste to the ground surface and evaluated the efficacy of carbonate chemistry in cementitious environment of the ILV for suppressing the volatilization of C-14. During the past year, a Special Analysis was prepared for Department of Energy approval to incorporate the results of these activities that increased the C-14 disposal limits for the ILV, thus allowing for disposal of the Reactor Moderator De-ionizers. Once the Special Analysis is approved by DOE, the actual disposal would be dependent on priority and funding, but the de-ionizers will be removed from the 'waste with no path to disposal list'. (authors)

  2. Determination of doses to different organs and prediction of health detriment, after hypothetical accident in mtr reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Amin, E A; Abd El-Ghani, A H [National Center of Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    As a result of pypothetical accidents with release of high amount of fission products, the doses to different organs consequent upon inhalation of radioactive fission products are calculated. The processes are modeled using the ORIGIN and TIRION-4 codes: source term, containment and activity enclosure, time dependent activity behaviour in the building, and radiation exposure in the reactor building. Prediction of health detriments were calculated using ICRP-60 nominal probability coefficients and organ doses determined for bone, lung, and thyroid gland, after whole body exposure from internal inhalation and external emmersion. 11 tabs.

  3. Some technical solutions on organization and technology of reactor room component mounting

    International Nuclear Information System (INIS)

    Romanovskij, V.I.

    1982-01-01

    Design of the production equipment for mounting sites of heat facilities of the Zaporozhe NPP is considered. Plan of the production equipment for mounting sites of heat facilities and flowsheet of mounting of supporting truss of the reactor are presented

  4. Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

    International Nuclear Information System (INIS)

    Zhu, Xiuping; Logan, Bruce E.

    2013-01-01

    Highlights: ► A new type of electro-Fenton system was developed for wastewater treatment. ► Degradation efficiency of organic pollutants was substantially improved. ► Operation cost was greatly reduced compared to other microbial fuel cell designs. -- Abstract: Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H 2 O 2 generation and Fe 2+ release for the Fenton reaction. In tests using phenol, 75 ± 2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22 h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants

  5. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density LEU fuels that are being developed by the RERTR program. High-density LEU dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits

  6. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density Leu fuels that are being developed by the Rarita program. High-density Leu dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits. (author)

  7. Effect of organic loading rate on dark fermentative hydrogen production in the continuous stirred tank reactor and continuous mixed immobilized sludge reactor from waste pastry hydrolysate.

    Science.gov (United States)

    Han, Wei; Hu, Yunyi; Li, Shiyi; Nie, Qiulin; Zhao, Hongting; Tang, Junhong

    2016-12-01

    Waste pastry (6%, w/v) was hydrolyzed by the produced glucoamylase and protease to obtain the glucose (19.8g/L) and free amino nitrogen (179mg/L) solution. Then, the effect of organic loading rate (OLR) (8-40kgCOD/(m 3 d)) on dark fermentative hydrogen production in the continuous stirred tank reactor (CSTR) and continuous mixed immobilized sludge reactor (CMISR) from waste pastry hydrolysate was investigated and compared. The maximum hydrogen production rate of CSTR (277.76mL/(hL)) and CMISR (320.2mL/(hL)) were achieved at OLR of 24kgCOD/(m 3 d) and 32kgCOD/(m 3 d), respectively. Carbon recovery ranged from 75.2-84.1% in the CSTR and CMISR with the balance assumed to be converted to biomass. One gram waste pastry could produce 0.33g (1.83mmol) glucose which could be further converted to 79.24mL (3.54mmol) hydrogen in the CMISR or 91.66mL (4.09mmol) hydrogen in the CSTR. This is the first study which reports dark fermentative hydrogen production from waste pastry. Copyright © 2016 Elsevier Ltd. All rights reserved.

  8. Effect of organic load on decolourization of textile wastewater containing acid dyes in upflow anaerobic sludge blanket reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijetunga, Somasiri, E-mail: swije2001@yahoo.com [Laboratory of Environmental Biotechnology, School of Biotechnology, Southern Yangtze University, 170 Huihe Road, Wuxi 214036 (China); Li Xiufen [Laboratory of Environmental Biotechnology, School of Biotechnology, Southern Yangtze University, 170 Huihe Road, Wuxi 214036 (China); Jian Chen, E-mail: jchen@sytu.edu.cn [Laboratory of Environmental Biotechnology, School of Biotechnology, Southern Yangtze University, 170 Huihe Road, Wuxi 214036 (China)

    2010-05-15

    Textile wastewater (TW) is one of the most hazardous wastewater for the environment when discharged without proper treatment. Biological treatment technologies have shown encouraging results over the treatment of recalcitrant compounds containing wastewaters. Upflow anaerobic sludge blanket reactor (UASB) was evaluated in terms of colour and the reduction of chemical oxygen demand (COD) with different organic loads using TW containing dyes belonging to different chemical groups. The study was performed using six different dye concentrations (10 mg/L, 25 mg/L, 50 mg/L, 100 mg/L, 150 mg/L, 300 mg/L) with three COD levels ({approx}1000 mg/L, {approx}2000 mg/L, {approx}3000 mg/L). Decolourization, COD removal and reactor stability were monitored. Over 85% of colour removal was observed with all dye concentrations with three organic loads. Acid Red 131 and Acid Yellow 79 were decolourized through biodegradation while Acid Blue 204 was decolourized due to adsorption onto anaerobic granules. COD removal was high in all dye concentrations, regardless of co-substrate levels. The reactor did not show any instability during the study. The activity of granules was not affected by the dyes. Methanothrix like bacteria were the dominant group in granules before introducing TW, however, they were reduced and cocci-shape microorganism increased after the treatment of textile wastewater.

  9. Effect of organic load on decolourization of textile wastewater containing acid dyes in upflow anaerobic sludge blanket reactor

    International Nuclear Information System (INIS)

    Wijetunga, Somasiri; Li Xiufen; Jian Chen

    2010-01-01

    Textile wastewater (TW) is one of the most hazardous wastewater for the environment when discharged without proper treatment. Biological treatment technologies have shown encouraging results over the treatment of recalcitrant compounds containing wastewaters. Upflow anaerobic sludge blanket reactor (UASB) was evaluated in terms of colour and the reduction of chemical oxygen demand (COD) with different organic loads using TW containing dyes belonging to different chemical groups. The study was performed using six different dye concentrations (10 mg/L, 25 mg/L, 50 mg/L, 100 mg/L, 150 mg/L, 300 mg/L) with three COD levels (∼1000 mg/L, ∼2000 mg/L, ∼3000 mg/L). Decolourization, COD removal and reactor stability were monitored. Over 85% of colour removal was observed with all dye concentrations with three organic loads. Acid Red 131 and Acid Yellow 79 were decolourized through biodegradation while Acid Blue 204 was decolourized due to adsorption onto anaerobic granules. COD removal was high in all dye concentrations, regardless of co-substrate levels. The reactor did not show any instability during the study. The activity of granules was not affected by the dyes. Methanothrix like bacteria were the dominant group in granules before introducing TW, however, they were reduced and cocci-shape microorganism increased after the treatment of textile wastewater.

  10. [Factors related to nurses' patient identification behavior and the moderating effect of person-organization value congruence climate within nursing units].

    Science.gov (United States)

    Kim, Young Mee; Kang, Seung Wan; Kim, Se Young

    2014-04-01

    This research was an empirical study designed to identify precursors and interaction effects related to nurses' patient identification behavior. A multilevel analysis methodology was used. A self-report survey was administered to registered nurses (RNs) of a university hospital in South Korea. Of the questionnaires, 1114 were analyzed. The individual-level factors that had a significantly positive association with patient identification behavior were person-organization value congruence, organizational commitment, occupational commitment, tenure at the hospital, and tenure at the unit. Significantly negative group-level precursors of patient identification behavior were burnout climate and the number of RNs. Two interaction effects of the person-organization value congruence climate were identified. The first was a group-level moderating effect in which the negative relationship between the number of RNs and patient identification behavior was weaker when the nursing unit's value congruence climate was high. The second was a cross-level moderating effect in which the positive relationship between tenure at the unit and patient identification behavior was weaker when value congruence climate was high. This study simultaneously tested both individual-level and group-level factors that potentially influence patient identification behavior and identified the moderating role of person-organization value congruence climate. Implications of these results are discussed.

  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. Performance of highly rated UO2 fuel in the WR-1 organic-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schankula, M. H.; Hastings, I. J.

    1977-07-15

    Information on oxide fuel behaviour in organic coolant was required as part of the organic-cooled power reactor (OCR) study. Of major interest were data on the release of fission gases from fuel operating at high fuel surface temperatures and low external restraint; features which are peculiar to the OCR. To provide these and other data, UO2 fuel with cold-worked Zr-2.5wt%Nb sheathing was irradiated in the WR-1 organic-cooled reactor to burnups of 135-154 MWh/kgU at a time-averaged linear power of 60-63 kW/m. Elements with 0.38 and 0.69 mm thick sheathing showed maximum diametral increases averaging 3.7 and 1.7% respectively at pellet mid-planes. Reduced fuel/sheath heat transfer resulting from a difference between internal gas pressure and coolant pressure produced high operating temperatures, and there was evidence of central melting in some elements. Fission gas releases were 30-60%. In the heat affected zone adjacent to brazed appendages, the diametral increases were lower, averaging 0.9 and 0.5% for 0.38 and 0.69 mm thick sheathing respectively. Heat treatment during the brazing process produced a local improvement in sheath creep strength. Highly rated oxide fuel irradiated in organic coolant will require sheathing with improved high temperature creep properties; heat-treated Zr-2.5 wt% Nb may provide this improvement.

  13. Using single-chamber microbial fuel cells as renewable power sources of electro-Fenton reactors for organic pollutant treatment

    KAUST Repository

    Zhu, Xiuping

    2013-05-01

    Electro-Fenton reactions can be very effective for organic pollutant degradation, but they typically require non-sustainable electrical power to produce hydrogen peroxide. Two-chamber microbial fuel cells (MFCs) have been proposed for pollutant treatment using Fenton-based reactions, but these types of MFCs have low power densities and require expensive membranes. Here, more efficient dual reactor systems were developed using a single-chamber MFC as a low-voltage power source to simultaneously accomplish H2O2 generation and Fe2+ release for the Fenton reaction. In tests using phenol, 75±2% of the total organic carbon (TOC) was removed in the electro-Fenton reactor in one cycle (22h), and phenol was completely degraded to simple and readily biodegradable organic acids. Compared to previously developed systems based on two-chamber MFCs, the degradation efficiency of organic pollutants was substantially improved. These results demonstrate that this system is an energy-efficient and cost-effective approach for industrial wastewater treatment of certain pollutants. © 2013 Elsevier B.V.

  14. Kinetics of concentration decay of specific organic matter in UASB reactors operating with and without return of aerobic sludge.

    Science.gov (United States)

    Pontes, P P; Chernicharo, C A L; Von Sperling, M

    2014-08-01

    This study aimed at assessing the influence of the return of excess aerobic sludge from a trickling filter (TF) upon the anaerobic digestion process in an upflow anaerobic sludge blanket (UASB) reactor, by evaluating its effect on the kinetics of the decay of specific organic matter (carbohydrates, proteins and lipids), as well as on the concentrations of volatile fatty acids in the UASB reactor. A pilot-scale UASB/TF system was used to perform the experiments, operating with (phase 2) and without (phase 1) excess sludge return from the TF to the UASB reactor. Sampling was carried out at different heights of the UASB reactor (0, 25, 125 and 225-cm height), and profile concentrations were determined for the following parameters: carbohydrates, proteins, lipids and volatile fatty acids. First-order kinetics showed the best fit to the decay of concentrations of carbohydrates, proteins, lipids and chemical oxygen demand (COD) in the UASB reactor. The parameters showing the best fit to the first-order kinetics were proteins and COD, during the sludge return phase. The occurrence of higher apparent reaction constants was further observed during the sludge return phase. For an influent COD concentration of 600 mg L-1 and hydraulic retention times of 2.1, 2.6 and 3.0 h in phase 1, the effluent COD concentrations were 125.3, 88.4 and 62.4 mg L-1, respectively, whereas in phase 2, the effluent COD concentrations were 75.5, 47.6 and 30.1 mg L-1, respectively.

  15. Experiences on reduction of reactor water silica and fresh resin leaching organics for Kuo-Sheng Nuclear Power Plant

    International Nuclear Information System (INIS)

    Wen, T-J.; Wang, C-H.

    2010-01-01

    The silica level in reactor water of Kuo-Sheng nuclear power plants has been slowly increased from 200 ppb to the high level above 500 ppb in recent years. The results obtained from steam/liquid mass balance calculation indicated that an increase of reactor water silica was mainly caused by continuing equilibrium leakage from deep bed condensate demineralizers, where the ion exchange zone was periodically disturbed by resin backwashing - scrubbing operation. The fastest and the most effective way to reduce the silica inventory in reactor system is to operate by continuously precoating of two sets of the reactor water clean up filter demineralizers to a lower effluent silica end point, and perhaps as frequently as three or four days. Leaching organic contaminants into feed water from the ion exchange resin becomes a key greater problem of current concern for the stable water quality promotion of condensate demineralizer. The presence of those impurities have practically been difficult to analyze by simple quality testing of the resin, and may result in as much as a hundred fold increase in chloride and sulfate in reactor water. As resin displacement with high leachable TOC, a repeated continuous soaking and effectively rinsing is required so that steady state TOC content less than 150 ppb should be achieved in an acceptably short period of time before put in-service. It is clear that cation resin containing high leachables generates high level of sulfates and sometimes also gives unexpected level of chlorides. The current TOC limits in condensate demineralizer effluent with 0.1 ppb become a significant experience to maintain reactor water soluble impurity in low levels. New resin should be subjected to TOC quality control testing prior to acceptance especially when first placed into service. TOC and organic chloride leachables for as-received virgin cation resin that are to be used in condensate polisher should be limited to be less than 100 mg-TOC and 0.5 mg-Cl per

  16. 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

  17. Contribution to the use of a solid moderator gas reactor, for naval propulsion; Contribution a l'etude d'un reacteur a gaz, a moderateur solide, pour propulsion navale

    Energy Technology Data Exchange (ETDEWEB)

    Pheline, J.; Gautier, A.

    1960-01-04

    In this contribution, the authors discuss works performed in France for the development of nuclear propulsion in merchant ships, notably for an oil tanker of 50.000 tons with 17 knot speed, i.e. a 20.000 Hp engine with an energy produced by a 60 MW gas reactor with a solid moderator and comprising 400 channels loaded with uranium oxide enriched ay 2.8 per cent and sheathed with a refractory alloy. The authors discuss the possible materials for the moderator, the heat transfer medium, the sheath, the fuel and the structures, and report technological studies (mechanical tests, irradiation tests) performed to investigate material properties and their behaviour in operation conditions. They report tests performed to investigate core structure characteristics with respect to neutrons. They finally briefly present a prototype.

  18. On the use of a composite moderator at the IBR-2 reactor: Advantages for the neutron-diffraction texture analysis of rocks

    Czech Academy of Sciences Publication Activity Database

    Bulavin, M. V.; Vasin, R.N.; Kulikov, S.A.; Lokajíček, Tomáš; Levin, D.M.

    2016-01-01

    Roč. 10, č. 4 (2016), s. 677-686 ISSN 1027-4510 Institutional support: RVO:67985831 Keywords : cold moderator * neutron-diffraction texture analysis * phase analysis * Rietveld method Subject RIV: DB - Geology ; Mineralogy

  19. The performance of a three-phase fluidized bed reactor in treatment of wastewater with high organic load

    Directory of Open Access Journals (Sweden)

    R. R. Souza

    2004-06-01

    Full Text Available An experimental study was carried out aiming to evaluate the performance of a three-phase fluidized bed bioreactor (FBBR used to treat milk wastewater. In this study three different concentrations of milk wastewater substrate (462, 825 and 1473 mg O2/L were tested. Using the same number of support particles, the results demonstrate that the average efficiency of COD removal decreased as the concentration of organic load in the substrate was increased. The growth of microorganism in the FBBR was followed by a count of viable cells in both liquid phase and the biofilms attached to the support. An increased number of viable cells were observed inside the reactor when it was used to degrade higher organic loads, with most of the cells on the support. The higher concentration of active biomass was responsible for achieving a relatively high absolute degradation of the wastewater containing the high organic load.

  20. Student-Fabricated Microfluidic Devices as Flow Reactors for Organic and Inorganic Synthesis

    Science.gov (United States)

    Feng, Z. Vivian; Edelman, Kate R.; Swanson, Benjamin P.

    2015-01-01

    Flow synthesis in microfluidic devices has been rapidly adapted in the pharmaceutical industry and in many research laboratories. Yet, the cost of commercial flow reactors is a major factor limiting the dissemination of this technology in the undergraduate curriculum. Here, we present a laboratory activity where students design and fabricate…

  1. EFFECT OF SULFATE LOADING RATE AND ORGANIC LOADING RATE ON ANAEROBIC BAFFLED REACTORS USED FOR TREATMENT OF SANITARY LANDFILL LEACHATES

    Directory of Open Access Journals (Sweden)

    O. Burbano-Figueroa

    2015-06-01

    Full Text Available AbstractThis study investigated the effect of organic loading rate (OLR and sulfate loading rate (SLR on landfill leachate treatment by a lab-scale anaerobic baffled reactor (ABR. Landfill leachate contained a concentration of organic matter between 3966 and 5090 mg COD.L-1 and no detectable amounts of sulfate. Reactors were started-up by feeding them with iron-sulfate at a SLR of 0.05 g SO42-.L-1.day-1 (4 weeks. Factorial design and response surface techniques were used to evaluate and optimize the effects of these operating variables on COD removal. ABRs were operated at OLRs ranging from 0.30 up to 6.84 g COD.L-1.day-1 by changes in influent volumetric flow. SO42- was added to the influent at a SRL from 0.06 to 0.13 g SO42-.L-1.day-1. The highest value of COD removal (66% was reached at an OLR of 3.58 g COD.L-1.day-1 and SLR of 0.09 g SO4-2.L-1.day-1 with a COD/SO4-2 ratio of 40. Under these conditions sulfate is mainly used for molecular hydrogen consumption while organic matter is preferentially degraded via methanogesis.

  2. 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

  3. Synergistic effect of plasmacatalyst and ozone in a pulsed corona discharge reactor on the decomposition of organic pollutants in water

    Energy Technology Data Exchange (ETDEWEB)

    Malik, Muhammad Arif [Applied Chemistry Laboratories, PINSTECH, PO Nilore, Islamabad (Pakistan)

    2003-11-01

    Plasmacatalytic effects of {alpha}-alumina, {gamma}-alumina, and silica gel in a pulsed corona discharge (PCD) reactor on the decomposition of aqueous methylene blue are described. Methylene blue concentration in the effluent was reduced to 23% of the inlet concentration by PCDs in water. Under the same experimental conditions, addition of {alpha}-alumina further reduced it to 8%, {gamma}-alumina to 4%, and silica gel to below the detection limits. PCDs with silica gel were run for >100 h in a continuous flow reactor and methylene blue in the effluent remained below the detection limit. A hybrid system of plasmacatalysis and ozonation is also described. Phenol concentration in the effluent was reduced to 84% of the inlet concentration by PCDs in water. Under the same experimental conditions, addition of either silica gel or ozone further reduced it to around 35%, and simultaneous addition of silica gel and ozone to 14% of inlet concentration. Decolourization of pre-adsorbed methylene blue on silica gel has been demonstrated. Adsorption and stabilization of the chemically active species on silica gel was indicated by experimental evidence. A significant improvement in the rate of decomposition of organic pollutants in water has been realized by hybridizing plasmacatalysis and ozonation in a PCD reactor.

  4. An examination of reliability critical items in liquid metal reactors: An analysis by the Centralized Reliability Data Organization (CREDO)

    International Nuclear Information System (INIS)

    Humphrys, B.L.; Haire, M.J.; Koger, K.H.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.; Okubo, Y.

    1987-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 21,000 components and approximately 1300 event records. A conservative estimation is that the total component operating hours is approaching 3.5 billion hours. Because data gathering for CREDO concentrates on event (failure) information, the work reported here focuses on the reliability information contained in CREDO and the development of reliability critical items lists. That is, components are ranked in prioritized lists from worst to best performers from a reliability standpoint. For the data contained in the CREDO data base, FFTF and JOYO show reliability growth; EBR-II reveals a slight unreliability growth for those components tracked by CREDO. However, tabulations of events which cause reactor shutdowns decrease with time at each site

  5. Synergistic effect of plasmacatalyst and ozone in a pulsed corona discharge reactor on the decomposition of organic pollutants in water

    International Nuclear Information System (INIS)

    Malik, Muhammad Arif

    2003-01-01

    Plasmacatalytic effects of α-alumina, γ-alumina, and silica gel in a pulsed corona discharge (PCD) reactor on the decomposition of aqueous methylene blue are described. Methylene blue concentration in the effluent was reduced to 23% of the inlet concentration by PCDs in water. Under the same experimental conditions, addition of α-alumina further reduced it to 8%, γ-alumina to 4%, and silica gel to below the detection limits. PCDs with silica gel were run for >100 h in a continuous flow reactor and methylene blue in the effluent remained below the detection limit. A hybrid system of plasmacatalysis and ozonation is also described. Phenol concentration in the effluent was reduced to 84% of the inlet concentration by PCDs in water. Under the same experimental conditions, addition of either silica gel or ozone further reduced it to around 35%, and simultaneous addition of silica gel and ozone to 14% of inlet concentration. Decolourization of pre-adsorbed methylene blue on silica gel has been demonstrated. Adsorption and stabilization of the chemically active species on silica gel was indicated by experimental evidence. A significant improvement in the rate of decomposition of organic pollutants in water has been realized by hybridizing plasmacatalysis and ozonation in a PCD reactor

  6. Effect of organic loading rate on methane and volatile fatty acids productions from anaerobic treatment of palm oil mill effluent in UASB and UFAF reactors

    Directory of Open Access Journals (Sweden)

    Sumate Chaiprapat

    2007-05-01

    Full Text Available Anaerobic treatment of palm oil mill effluent (POME with the separation of the acidogenic and methanogenic phase was studied in an up-flow anaerobic sludge blanket (UASB reactor and an up-flowanaerobic filter (UFAF reactor. Furthermore, the effect of OLR on methane and volatile fatty acid productions in UASB and UFAF reactors was investigated. In this research, UASB as acidogenic reactor wasused for volatile fatty acid production and UFAF as methanogenic reactor was used for methane production. Therefore, POME without pH adjustment was used as influent for the UASB reactor. Moreover, the syntheticwastewater with pH adjustment to 6.00 was fed into the UFAF reactor. The inoculum source for both reactors was the combination of POME sludge collected from the CSTR of a POME treatment plant and granulesludge collected from the UASB reactor of a frozen sea food industry treatment plant. During experimental operation, the organic loading rate (OLR was gradually increased from 2.50 to 17.5 g COD/l/day in theUASB reactor and 1.10 to 10.0 g COD/l/day in the UFAF reactor. Consequently, hydraulic retention time (HRT ranged from 20.0 to 2.90 days in the UASB reactor and from 13.5 to 1.50 days in the UFAF reactor.The result showed that the COD removal efficiency from both reactors was greater than 60.0%. In addition, the total volatile fatty acids increased with the increasing OLR. The total volatile fatty acids and acetic acidproduction in the UASB reactor reached 5.50 g/l and 4.90 g/l, respectively at OLR of 17.5 g COD/l/day and HRT of 2.90 days before washout was observed. In the UFAF reactor, the methane and biogas productionincreased with increasing OLR until an OLR of 7.50 g COD/l/day. However, the methane and biogas production significantly decreased when OLR increased up to 10.0 g COD/l/day. Therefore, the optimum OLR inthe laboratory-scale UASB and UFAF reactors were concluded to be 15.5 and 7.50 g COD/l/day, respectively.

  7. Organization structure as a moderator of the relationship between procedural justice, interactional justice, perceived organizational support, and supervisory trust.

    Science.gov (United States)

    Ambrose, Maureen L; Schminke, Marshall

    2003-04-01

    Organizational justice researchers recognize the important role organization context plays in justice perceptions, yet few studies systematically examine contextual variables. This article examines how 1 aspect of context--organizational structure--affects the relationship between justice perceptions and 2 types of social exchange relationships, organizational and supervisory. The authors suggest that under different structural conditions, procedural and interactional justice will play differentially important roles in determining the quality of organizational social exchange (as evidenced by perceived organizational support [POS]) and supervisory social exchange (as evidenced by supervisory trust). In particular, the authors hypothesized that the relationship between procedural justice and POS would be stronger in mechanistic organizations and that the relationship between interactional justice and supervisory trust would be stronger in organic organizations. The authors' results support these hypotheses.

  8. Identification of tertiary butyl alcohol (TBA)-utilizing organisms in BioGAC reactors using 13C-DNA stable isotope probing.

    Science.gov (United States)

    Aslett, Denise; Haas, Joseph; Hyman, Michael

    2011-09-01

    Biodegradation of the gasoline oxygenates methyl tertiary-butyl ether (MTBE) and ethyl tertiary-butyl ether (ETBE) can cause tertiary butyl alcohol (TBA) to accumulate in gasoline-impacted environments. One remediation option for TBA-contaminated groundwater involves oxygenated granulated activated carbon (GAC) reactors that have been self-inoculated by indigenous TBA-degrading microorganisms in ground water extracted from contaminated aquifers. Identification of these organisms is important for understanding the range of TBA-metabolizing organisms in nature and for determining whether self-inoculation of similar reactors is likely to occur at other sites. In this study (13)C-DNA-stable isotope probing (SIP) was used to identify TBA-utilizing organisms in samples of self-inoculated BioGAC reactors operated at sites in New York and California. Based on 16S rRNA nucleotide sequences, all TBA-utilizing organisms identified were members of the Burkholderiales order of the β-proteobacteria. Organisms similar to Cupriavidus and Methylibium were observed in both reactor samples while organisms similar to Polaromonas and Rhodoferax were unique to the reactor sample from New York. Organisms similar to Hydrogenophaga and Paucibacter strains were only detected in the reactor sample from California. We also analyzed our samples for the presence of several genes previously implicated in TBA oxidation by pure cultures of bacteria. Genes Mpe_B0532, B0541, B0555, and B0561 were all detected in (13)C-metagenomic DNA from both reactors and deduced amino acid sequences suggested these genes all encode highly conserved enzymes. One gene (Mpe_B0555) encodes a putative phthalate dioxygenase-like enzyme that may be particularly appropriate for determining the potential for TBA oxidation in contaminated environmental samples.

  9. Graphite moderated 252Cf source

    International Nuclear Information System (INIS)

    Sajo B, L.; Barros, H.; Greaves, E. D.; Vega C, H. R.

    2014-08-01

    The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a 252 Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252 Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

  10. Biomass characteristics in three sequencing batch reactors treating a wastewater containing synthetic organic chemicals

    DEFF Research Database (Denmark)

    Hu, Z.Q.; Ferraina, R.A.; Ericson, J.F.

    2005-01-01

    in all reactors. In contrast, effluent 3-nitrobenzoate was recorded when its influent concentration was increased to 5 mg L-1 and dropped only to below 1 mg L-1 after 300 days of operation. The competent (active) biomass fractions for these compounds were between 0.04% and 5.52% of the total biomass...... characteristics in the aerobic SBR and SBBR. While all reactors had very good COD removal (> 90%) and displayed nitrification, substantial nitrogen removal (74%) was only achieved in the anoxic/aerobic SBR. During the entire operational period, benzoate, theophylline and 4-chlorophenol were completely removed...... inferred from substrate-specific microbial enumerations. The measured competent biomass fractions for 4-chlorophenol and 3-nitrobenzoate degradation were significantly lower than the influent COD fractions of these compounds. Correspondent to the highest competent biomass fraction for benzoate degradation...

  11. Organization and administrative procedures for the control of the PUSPATI research reactor

    International Nuclear Information System (INIS)

    Nahrul Khair Rashid.

    1983-01-01

    The organizational set-up of PUSPATI, the administration and operation of reactor are discussed in detail. The nuclear related activities in Malaysia are governed by Radioactive Substances Act, 1968. An internal PUSPATI Safety Committee was formed to operate in a much similar manner to a typical nuclear regulatory body but in a smaller state enough to cater the present needs. Most of the rules and regulations were adopted from several guides, technical report, series of IAEA and the practice of other similar facilities. For inspection, the reports contain data on the operation, usage, maintenance and modifications, made on the reactor are submitted monthly. From these reports, a quarterly report will be produced and submitted to the PUSPATI Safety Committee. (A.J.)

  12. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1981-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drivemechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displayer rods through the reactor vessel

  13. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Sherwood, D.G.; Wilson, J.F.; Salton, R.B.; Fensterer, H.F.

    1982-01-01

    A mechanical spectral shift reactor comprises apparatus for inserting and withdrawing water displacer elements from the reactor core for selectively changing the water-moderator volume in the core thereby changing the reactivity of the core. The apparatus includes drive mechanisms for moving the displacer elements relative to the core and guide mechanisms for guiding the displacer rods through the reactor vessel. (author)

  14. How Environmental Uncertainty Moderates the Effect of Relative Advantage and Perceived Credibility on the Adoption of Mobile Health Services by Chinese Organizations in the Big Data Era.

    Science.gov (United States)

    Chen, Xing; Zhang, Xing

    2016-01-01

    Despite the importance of adoption of mobile health services by an organization on the diffusion of mobile technology in the big data era, it has received minimal attention in literature. This study investigates how relative advantage and perceived credibility affect an organization's adoption of mobile health services, as well as how environmental uncertainty changes the relationship of relative advantage and perceived credibility with adoption. A research model that integrates relative advantage, perceived credibility, environmental uncertainty, and an organization's intention to use mobile health service is developed. Quantitative data are collected from senior managers and information systems managers in 320 Chinese healthcare organizations. The empirical findings show that while relative advantage and perceived credibility both have positive effects on an organization's intention to use mobile health services, relative advantage plays a more important role than perceived credibility. Moreover, environmental uncertainty positively moderates the effect of relative advantage on an organization's adoption of mobile health services. Thus, mobile health services in environments characterized with high levels of uncertainty are more likely to be adopted because of relative advantage than in environments with low levels of uncertainty.

  15. Dynamic bioconversion mathematical modelling and simulation of urban organic waste co-digestion in continuously stirred tank reactor

    DEFF Research Database (Denmark)

    Fitamo, Temesgen Mathewos; Boldrin, Alessio; Dorini, G.

    of this study was to apply a dynamic mathematical model to simulate the co-digestion of different urban organic wastes (UOW). The modelling was based on experimental activities, during which two reactors (R1, R2) were operated at hydraulic retention times (HRT) of 30, 20, 15, 10 days, in thermophilic conditions......The application of anaerobic digestion (AD) as process technology is increasing worldwide: the production of biogas, a versatile form of renewable energy, from biomass and organic waste materials allows mitigating greenhouse gas emission from the energy and transportation sectors while treating...... waste. However, the successful operation of AD processes is challenged by economic and technological issues. To overcome these barriers, mathematical modelling of the bioconversion process can provide support to develop strategies for controlling and optimizing the AD process. The objective...

  16. Effect of granular activated carbon concentration on the content of organic matter and salt, influencing E. coli activity and survival in fluidized bed disinfection reactor

    NARCIS (Netherlands)

    Racyte, J.; Langenhoff, A.A.M.; Ribeiro, A.F.M.M.R.; Paulitsch-Fuchs, A.H.; Bruning, H.; Rijnaarts, H.

    2014-01-01

    Granular activated carbon (GAC) is used in water treatment systems, typically to remove pollutants such as natural organic matter, volatile organic compounds, chlorine, taste, and odor. GAC is also used as a key component of a new technology that combines a fluidized bed reactor with radio frequency

  17. Denitrifying capability and community dynamics of glycogen accumulating organisms during sludge granulation in an anaerobic-aerobic sequencing batch reactor

    Science.gov (United States)

    Bin, Zhang; Bin, Xue; Zhigang, Qiu; Zhiqiang, Chen; Junwen, Li; Taishi, Gong; Wenci, Zou; Jingfeng, Wang

    2015-08-01

    Denitrifying capability of glycogen accumulating organisms (GAOs) has received great attention in environmental science and microbial ecology. Combining this ability with granule processes would be an interesting attempt. Here, a laboratory-scale sequencing batch reactor (SBR) was operated to enrich GAOs and enable sludge granulation. The results showed that the GAO granules were cultivated successfully and the granules had denitrifying capability. The batch experiments demonstrated that all NO3--N could be removed or reduced, some amount of NO2--N were accumulated in the reactor, and N2 was the main gaseous product. SEM analysis suggested that the granules were tightly packed with a large amount of tetrad-forming organisms (TFOs); filamentous bacteria served as the supporting structures for the granules. The microbial community structure of GAO granules was differed substantially from the inoculant conventional activated sludge. Most of the bacteria in the seed sludge grouped with members of Proteobacterium. FISH analysis confirmed that GAOs were the predominant members in the granules and were distributed evenly throughout the granular space. In contrast, PAOs were severely inhibited. Overall, cultivation of the GAO granules and utilizing their denitrifying capability can provide us with a new approach of nitrogen removal and saving more energy.

  18. Moderator Configuration Options for ESS

    DEFF Research Database (Denmark)

    Zanini, L.; Batkov, K.; Klinkby, Esben Bryndt

    2016-01-01

    The current, still evolving status of the design and the optimization work for the moderator configuration for the European Spallation Source is described. The moderator design has been strongly driven by the low-dimensional moderator concept recently proposed for use in spallation neutron sources...... or reactors. Quasi-two dimensional, disc- or tube-shaped moderators,can provide strong brightness increase (factor of 3 or more) with respect to volume para-H2moderators, which constitute the reference, state-of-the-art technology for high-intensity coupled moderators. In the design process other, more...... conventional, principles were also considered,such as the importance of moderator positioning, of the premoderator, and beam extraction considerations. Different design and configuration options are evaluated and compared with the reference volume moderator configuration described in the ESS Technical Design...

  19. Individual and combined effects of organic, toxic, and hydraulic shocks on sequencing batch reactor in treating petroleum refinery wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Mizzouri, Nashwan Sh., E-mail: nashwan_mizzouri@yahoo.com [Department of Civil Engineering, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia); Department of Civil Engineering, University of Duhok, Kurdistan (Iraq); Shaaban, Md Ghazaly [Department of Civil Engineering, University of Malaya, Lembah Pantai, 50603 Kuala Lumpur (Malaysia)

    2013-04-15

    Highlights: ► This research focuses on the combined impact of shock loads on the PRWW treatment. ► System failure resulted when combined shock of organic and hydraulic was applied. ► Recovery was achieved by replacing glucose with PRWW and OLR was decreased to half. ► Worst COD removals were 68.9, and 57.8% for organic, and combined shocks. -- Abstract: This study analyzes the effects of toxic, hydraulic, and organic shocks on the performance of a lab-scale sequencing batch reactor (SBR) with a capacity of 5 L. Petroleum refinery wastewater (PRWW) was treated with an organic loading rate (OLR) of approximately 0.3 kg chemical oxygen demand (COD)/kg MLSS d at 12.8 h hydraulic retention time (HRT). A considerable variation in the COD was observed for organic, toxic, hydraulic, and combined shocks, and the worst values observed were 68.9, 77.1, 70.2, and 57.8%, respectively. Improved control of toxic shock loads of 10 and 20 mg/L of chromium (VI) was identified. The system was adversely affected by the organic shock when a shock load thrice the normal value was used, and this behavior was repeated when the hydraulic shock was 4.8 h HRT. The empirical recovery period was greater than the theoretical period because of the inhibitory effects of phenols, sulfides, high oil, and grease in the PRWW. The system recovery rates from the shocks were in the following order: toxic, organic, hydraulic, and combined shocks. System failure occurred when the combined shocks of organic and hydraulic were applied. The system was resumed by replacing the PRWW with glucose, and the OLR was reduced to half its initial value.

  20. 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.)

  1. Chromophoric Dissolved Organic Matter and Dissolved Organic Carbon from Sea-Viewing Wide Field-of-View Sensor (SeaWiFS), Moderate Resolution Imaging Spectroradiometer (MODIS) and MERIS Sensors: Case Study for the Northern Gulf of Mexico

    OpenAIRE

    Blake A. Schaeffer; Thomas S. Bianchi; Eurico J. D'Sa; Christopher L. Osburn; Nazanin Chaichi Tehrani

    2013-01-01

    Empirical band ratio algorithms for the estimation of colored dissolved organic matter (CDOM) and dissolved organic carbon (DOC) for Sea-viewing Wide Field-of-view Sensor (SeaWiFS), Moderate Resolution Imaging Spectroradiometer (MODIS) and MERIS ocean color sensors were assessed and developed for the northern Gulf of Mexico. Match-ups between in situ measurements of CDOM absorption coefficients at 412 nm (aCDOM(412)) with that derived from SeaWiFS were examined using two previously reported r...

  2. 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.

  3. A comparison of predicted and measured graphite moderator behaviour during 16 years' operation of the Windscale advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    Prince, N.

    1980-01-01

    The Windscale AGR has operated since January 1963 at a cumulative load factor of nearly 70% during which time the peak irradiation damage dose has built up to more than 5x10 n/cm 2 (equivalent DIDO nickel), well beyond its original design life. This paper recounts the findings of monitoring measurements on the moderator at high exposure levels with regard to radiolytic oxidation and various aspects of dimensional change behaviour. It is shown that the measured dimensional changes are in good agreement with predictions based on small specimens irradiated in MTR's, thus confirming the absence of any size effect and adding confidence to predictive methods. However, recent measurements of channel straightness show that the observed distortions are only about 10% of the maximum predictions, perhaps due to localised creep at the brick ends creating flats which impart some stability to the columns of moderator bricks. The magnitude of radiolytic oxidation determined by trepanning specimens from the core in 1976 was found to be only about 5%, whereas it was thought possible that peak weight losses would conceivably be as high as 11% due to the depleted concentration of methane inhibitor reaching the brick interior by diffusion processes. It has subsequently been shown by calculation that this result is consistent with the existence of radial pressure drops across the moderator brick walls giving greater penetration of methane inhibitor. (author)

  4. Influence of nutrients on biomass evolution in an upflow anaerobic sludge blanket reactor degrading sulfate-laden organics.

    Science.gov (United States)

    Patidar, S K; Tare, Vinod

    2004-01-01

    This paper describes the effect of the nutrients iron (Fe), nickel (Ni), zinc (Zn), cobalt (Co), and molybdenum (Mo) on biomass evolution in an upflow anaerobic sludge blanket (UASB) reactor metabolizing synthetic sulfate-laden organics at varying operating conditions during a period of 540 days. A bench-scale model of a UASB reactor was operated at a temperature of 35 degrees C for a chemical oxygen demand-to-sulfate (COD/SO4(2-)) ratio of 8.59 to 2.0, a sulfate loading rate of 0.54 to 1.88 kg SO4(2-)/m3 x d, and an organic loading rate of 1.9 to 5.75 kg COD/m3 x d. Biomass was characterized in terms of total methanogenic activity, acetate-utilizing methanogenic activity, total sulfidogenic activity, acetate-utilizing sulfidogenic activity, and scanning electron microscopy (SEM). Nickel and cobalt limitation appears to affect the activity of hydrogen-utilizing methane-producing bacteria (HMPB) significantly without having an appreciable effect on the activity of acetate-utilizing methane-producing bacteria (AMPB). Nickel and cobalt supplementation resulted in increased availability and, consequently, restoration of biomass activity and process performance. Iron limitation and sulfidogenic conditions resulted in the growth of low-density, hollow, fragile granules that washed out, causing process instability and performance deterioration. Iron and cobalt supplementation indicated significant stimulation of AMPB with slight inhibition of HMPB. Examination of biomass through SEM indicated a population shift with dominance of sarcina-type organisms and the formation of hollow granules. Granule disintegration was observed toward the end of the study.

  5. Performance of sulphate- and selenium-reducing biochemical reactors using different ratios of labile to recalcitrant organic materials.

    Science.gov (United States)

    Mirjafari, Parissa; Baldwin, Susan A

    2015-01-01

    Successful operation of sulphate-reducing bioreactors using complex organic materials depends on providing a balance between more easily degrading material that achieves reasonable kinetics and low hydraulic retention times, and more slowly decomposing material that sustains performance in the long term. In this study, two organic mixtures containing the same ingredients typical of bioreactors used at mine sites (woodchips, hay and cow manure) but with different ratios of wood (recalcitrant) to hay (labile) were tested in six continuous flow bioreactors treating synthetic mine-affected water containing 600 mg/L of sulphate and 15 μg/L of selenium. The reactors were operated for short (5-6 months) and long (435-450 days) periods of time at the same hydraulic retention time of 15 days. There were no differences in the performance of the bioreactors in terms of sulphate-reduction over the short term, but the wood-rich bioreactors experienced variable and sometimes unreliable sulphate-reduction over the long term. Presence of more hay in the organic mixture was able to better sustain reliable performance. Production of dissolved organic compounds due to biodegradation within the bioreactors was detected for the first 175-230 days, after which their depletion coincided with a crash phase observed in the wood-rich bioreactors only.

  6. Moderation of neutron energy

    International Nuclear Information System (INIS)

    Marlatt, G.R.

    1986-01-01

    This patent describes a nuclear reactor system having a nuclear reactor which has a core including fuel assemblies, means for transmitting through the core a coolant, the coolant having a predetermined neutron-energy moderating property, sealed tubes in the core, each tube containing a material having a different neutron-energy moderating property than the coolant, means, when actuated, to engage at least certain of the tubes, for opening certain of the tubes to permit the coolant to replace the material in the tubes thereby to change the energy spectrum of the neutrons in the reactor, hydraulic means, connected to the opening means, for actuating the opening means to engage certain of the tubes to open the tubes. A device, external to the reactor, connected to the hydraulic means controlls the actuation of the opening means, the opening means being so set with reference to the tubes that only certain of the tubes are opened at any time as the opening means is advanced towards the tubes by the hydraulic means

  7. 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)

  8. Separated type nuclear superheating reactor

    International Nuclear Information System (INIS)

    Hida, Kazuki.

    1993-01-01

    In a separated type nuclear superheating reactor, fuel assemblies used in a reactor core comprise fuel rods made of nuclear fuel materials and moderator rods made of solid moderating materials such as hydrogenated zirconium. Since the moderating rods are fixed or made detachable, high energy neutrons generated from the fuel rods are moderated by the moderating rods to promote fission reaction of the fuel rods. Saturated steams supplied from the BWR type reactor by the fission energy are converted to high temperature superheated steams while passing through a steam channel disposed between the fuel rods and the moderating rods and supplied to a turbine. Since water is not used but solid moderating materials sealed in a cladding tube are used as moderation materials, isolation between superheated steams and water as moderators is not necessary. Further, since leakage of heat is reduced to improve a heat efficiency, the structure of the reactor core is simplified and fuel exchange is facilitated. (N.H.)

  9. 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.)

  10. High Flux Isotope Reactor (HFIR)

    Data.gov (United States)

    Federal Laboratory Consortium — The HFIR at Oak Ridge National Laboratory is a light-water cooled and moderated reactor that is the United States’ highest flux reactor-based neutron source. HFIR...

  11. A side-by-side comparison of two systems of sequencing coupled reactors for anaerobic digestion of the organic fraction of municipal solid waste.

    Science.gov (United States)

    Poggi-Varaldo, Héctor M; Alzate-Gaviria, Liliana M; Pérez-Hernández, Antonino; Nevarez-Morillón, Virginia G; Rinderknecht-Seijas, Noemí

    2005-06-01

    The objective of this work was to compare the performance of two laboratory-scale, mesophilic systems aiming at the anaerobic digestion of the organic fraction of municipal solid wastes (OFMSW). The first system consisted of two coupled reactors packed with OFMSW (PBR1.1-PBR1.2) and the second system consisted of an upflow anaerobic sludge bed reactor (UASB) coupled to a packed reactor (UASB2.1-PBR2.2). For the start-up phase, both reactors PBR 1.1 and the UASB 2.1 (also called leading reactors) were inoculated with a mixture of non-anaerobic inocula and worked with leachate and effluent full recirculation, respectively. Once a full methanogenic regime was achieved in the leading reactors, their effluents were fed to the fresh-packed reactors PBR1.2 and PBR2.2, respectively. The leading PBR 1.1 reached its full methanogenic regime after 118 days (Tm, time to achieve methanogenesis) whereas the other leading UASB 2.1 reactor reached its full methanogenesis regime after only 34 days. After coupling the leading reactors to the corresponding packed reactors, it was found that both coupled anaerobic systems showed similar performances regarding the degradation of the OFMSW. Removal efficiencies of volatile solids and cellulose and the methane pseudo-yield were 85.95%, 80.88% and 0.109 NL CH4 g(-1) VS(fed) in the PBR-PBR system; and 88.75%, 82.61% and 0.115 NL CH4 g(-1) VS(fed0 in the UASB-PBR system [NL, normalized litre (273 degrees K, 1 ata basis)]. Yet, the second system UASB-PBR system showed a faster overall start-up.

  12. Effect of organic loading rate on anaerobic treatment of slaughterhouse wastewater in a fluidised-bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Borja, R. [Consejo Superior de Investigaciones Cientificas, Seville (Spain). Inst. de la Grasa; Banks, C.J.; Zhengjian Wang [Manchester Univ. (United Kingdom). Inst. of Science and Technology

    1995-09-01

    COD removal efficiencies in the range 75.0-98.9% were achieved in an aerobic fluidised-bed reactor treating slaughterhouse wastewater, when evaluated at organic loading rates (OLR) of between 2.9 and 54.0 g COD/l.d, hydraulic retention times (HRT) of between 0.5 and 8 h and feed COD concentrations of between 250 and 4500 mg/l. More than 94% of feed COD could be removed up to OLR of about 27 g COD/l.d. Up to 0.320 litres of methane were produced per gram of COD removed and this methane production rate was independent of the OLR applied in this investigation. Volatile fatty acid (VFA) concentration in the reactor increased sharply at an OLR of about 30 g COD/l.d and, therefore, sufficient alkalinity should be provided to prevent pH from dropping to an undesirable level. The anaerobic fluidised-bed system can be operated at a significantly higher liquid throughput than other previously reported systems while maintaining its excellent efficiency. (Author)

  13. Cascade degradation of organic matters in brewery wastewater using a continuous stirred microbial electrochemical reactor and analysis of microbial communities

    Science.gov (United States)

    Wang, Haiman; Qu, Youpeng; Li, Da; Ambuchi, John J.; He, Weihua; Zhou, Xiangtong; Liu, Jia; Feng, Yujie

    2016-01-01

    A continuous stirred microbial electrochemical reactor (CSMER), comprising of a complete mixing zone (CMZ) and microbial electrochemical zone (MEZ), was used for brewery wastewater treatment. The system realized 75.4 ± 5.7% of TCOD and 64.9 ± 4.9% of TSS when fed with brewery wastewater concomitantly achieving an average maximum power density of 304 ± 31 m W m−2. Cascade utilization of organic matters made the CSMER remove a wider range of substrates compared with a continuous stirred tank reactor (CSTR), in which process 79.1 ± 5.6% of soluble protein and 86.6 ± 2.2% of soluble carbohydrates were degraded by anaerobic digestion in the CMZ and short-chain volatile fatty acids were further decomposed and generated current in the MEZ. Co-existence of fermentative bacteria (Clostridium and Bacteroides, 19.7% and 5.0%), acetogenic bacteria (Syntrophobacter, 20.8%), methanogenic archaea (Methanosaeta and Methanobacterium, 40.3% and 38.4%) and exoelectrogens (Geobacter, 12.4%) as well as a clear spatial distribution and syntrophic interaction among them contributed to the cascade degradation process in CSMER. The CSMER shows great promise for practical wastewater treatment application due to high pre-hydrolysis and acidification rate, high energy recovery and low capital cost. PMID:27270788

  14. Determination of mean molecular weights in organic reactor coolants. III. Differential cryoscopy with thermoelectric thermometer

    International Nuclear Information System (INIS)

    Becerro, E.; Carreira, M.

    1968-01-01

    The solubility problems raised by some components of the polymeric residue of irradiated polyphenolic coolants, which make it necessary to operate with very small samples, have been solved by means of a differential cryoscopic technique using a thermoelectric thermometer (thermal) as sensitive element. The method is based on the direct measurement of the difference between the freezing points of the investigated solution and of a reference solution whose concentration may be changed at will. The change of Δ V (mV) versus c(molal) is linear, the equivalent point being determined either analytically or graphically depending on the required accuracy. The method has been tested by measurements on pure polyphenyls, using diphenyl ether as solvent. It has been also applied to the main prospective coolants for the DON reactor. Working with 10 2 molal solutions the accuracy is better than ± 2 per cent. (Author) 2 refs

  15. Trade study for kWe class space reactors

    Science.gov (United States)

    Bost, Donald S.

    Recent interest by NASA and other government agencies in space reactor power systems with power levels in the 1 to 100 kWe range has prompted a review of earlier space reactor programs, as well as the ongoing SP-100 program, to identify a system that will best fulfill their needs. The candidate reactor types that were reviewed are listed. They are categorized according to the method of heat removal. The five types are: conduction cooled, heat pipe cooled, liquid metal cooled, in-core thermionic and gas cooled. The UZrH moderated reactor coupled with an organic Rankine cycle power conversion system provides an attractive system for multikilowatt, long lived missions. The reactor requires a minimum development because a similar reactor has already flown and the ORC is being developed for use in the Dynamic Isotope Power System (DIPS) and on the Space Station.

  16. Fe(II)-regulated moderate pre-oxidation of Microcystis aeruginosa and formation of size-controlled algae flocs for efficient flotation of algae cell and organic matter.

    Science.gov (United States)

    Qi, Jing; Lan, Huachun; Liu, Ruiping; Liu, Huijuan; Qu, Jiuhui

    2018-06-15

    The coagulation/flocculation/flotation (C/F/F) process is becoming a popular method for algae-laden water treatment. However, the efficiency of flotation is highly dependent on the ability of the preceding coagulation/flocculation process to form flocculated algae flocs. This study aims to improve the Microcystis aeruginosa flotation efficiency from algae cell and organic matter aspects by applying Fe(II)-regulated pretreatment enhanced Al coagulation process. The ability of the C/F/F process to remove cyanobacterial cells can be enhanced from 8% to 99% at a Fe(II) dose of 30 μM. The Al dose needed can be reduced by more than half while achieving successful flotation. The introduced Fe(II) after KMnO 4 can not only realize moderate pre-oxidation of cyanobacterial cells, but also form in-situ Fe(III). The DOC value can also be decreased significantly due to the formation of in-situ Fe(III), which is more efficient in dissolved organic matter (DOM) removal compared with pre-formed Fe(III). In addition, the gradually hydrolyzed in-situ Fe(III) can facilitate the hydrolysis of Al as a dual-coagulant and promote the clustering and cross-linking of Al hydrolyzates, which can enhance the formation of size-controlled algae flocs. Finally, the size-controlled algae flocs can be effectively floated by the bubbles released in the flotation process due to the efficient collision and attachment between flocs and bubbles. Therefore, the efficient flotation of algae cell and organic matter can be realized by the Fe(II) regulated moderate pre-oxidation of M. aeruginosa and formation of size-controlled algae flocs. Copyright © 2018 Elsevier Ltd. All rights reserved.

  17. Highly efficient phosphor-converted white organic light-emitting diodes with moderate microcavity and light-recycling filters.

    Science.gov (United States)

    Cho, Sang-Hwan; Oh, Jeong Rok; Park, Hoo Keun; Kim, Hyoung Kun; Lee, Yong-Hee; Lee, Jae-Gab; Do, Young Rag

    2010-01-18

    We demonstrate the combined effects of a microcavity structure and light-recycling filters (LRFs) on the forward electrical efficiency of phosphor-converted white organic light-emitting diodes (pc-WOLEDs). The introduction of a single pair of low- and high-index layers (SiO(2)/TiO(2)) improves the blue emission from blue OLED and the insertion of blue-passing and yellow-reflecting LRFs enhances the forward yellow emission from the YAG:Ce(3+) phosphors layers. The enhancement of the luminous efficacy of the forward white emission is 1.92 times that of a conventional pc-WOLED with color coordinates of (0.34, 0.34) and a correlated color temperature of about 4800 K.

  18. A physiologically based toxicokinetic (PBTK) model for moderately hydrophobic organic chemicals in the European eel (Anguilla anguilla)

    Energy Technology Data Exchange (ETDEWEB)

    Brinkmann, Markus [Department of Ecosystem Analysis, Institute for Environmental Research, ABBt — Aachen Biology and Biotechnology, RWTH Aachen University, Aachen (Germany); Freese, Marko; Pohlmann, Jan-Dag; Kammann, Ulrike [Thünen Institute of Fisheries Ecology, Hamburg (Germany); Preuss, Thomas G. [Environmental Biology and Chemodynamics, Institute for Environmental Research, ABBt — Aachen Biology and Biotechnology, RWTH Aachen University, Aachen (Germany); Buchinger, Sebastian; Reifferscheid, Georg [Federal Institute of Hydrology (BFG), Department G3: Biochemistry, Ecotoxicology, Koblenz (Germany); Beiermeister, Anne; Hanel, Reinhold [Thünen Institute of Fisheries Ecology, Hamburg (Germany); Hollert, Henner, E-mail: Henner.hollert@bio5.rwth-aachen.de [Department of Ecosystem Analysis, Institute for Environmental Research, ABBt — Aachen Biology and Biotechnology, RWTH Aachen University, Aachen (Germany); State Key Laboratory of Pollution Control and Resource Reuse, School of the Environment, Nanjing University, Nanjing (China); College of Resources and Environmental Science, Chongqing University, Chongqing (China); Key Laboratory of Yangtze Water Environment, Ministry of Education, Tongji University, Shanghai 200092 (China)

    2015-12-01

    The European eel (Anguilla anguilla) is a facultatively catadromous fish species with a complex life cycle. Its current population status is alarming: recruitment has decreased drastically since the 1980s and its stock is still considered to be outside safe biological limits. Although there is no consensus on the reasons for this situation, it is currently thought to have resulted from a combination of different stressors, including anthropogenic contaminants. To deepen our understanding of the processes leading to the accumulation of lipophilic organic contaminants in yellow eels (i.e. the feeding, continental growth stage), we developed a physiologically based toxicokinetic model using our own data and values from the literature. Such models can predict the uptake and distribution of water-borne organic chemicals in the whole fish and in different tissues at any time during exposure. The predictive power of the model was tested against experimental data for six chemicals with n-octanol-water partitioning coefficient (log K{sub ow}) values ranging from 2.13–4.29. Model performance was excellent, with a root mean squared error of 0.28 log units. This model has the potential to help identify suitable habitats for restocking under eel management plans. - Highlights: • A PBTK model was developed for European eel (Anguilla anguilla). • Own experimental data and data from the literature were used for parameterization. • The predictive power of the model was excellent, with RMSE of 0.28 log units. • The developed model can be amended with sub-models for dietary and dermal exposure.

  19. A physiologically based toxicokinetic (PBTK) model for moderately hydrophobic organic chemicals in the European eel (Anguilla anguilla)

    International Nuclear Information System (INIS)

    Brinkmann, Markus; Freese, Marko; Pohlmann, Jan-Dag; Kammann, Ulrike; Preuss, Thomas G.; Buchinger, Sebastian; Reifferscheid, Georg; Beiermeister, Anne; Hanel, Reinhold; Hollert, Henner

    2015-01-01

    The European eel (Anguilla anguilla) is a facultatively catadromous fish species with a complex life cycle. Its current population status is alarming: recruitment has decreased drastically since the 1980s and its stock is still considered to be outside safe biological limits. Although there is no consensus on the reasons for this situation, it is currently thought to have resulted from a combination of different stressors, including anthropogenic contaminants. To deepen our understanding of the processes leading to the accumulation of lipophilic organic contaminants in yellow eels (i.e. the feeding, continental growth stage), we developed a physiologically based toxicokinetic model using our own data and values from the literature. Such models can predict the uptake and distribution of water-borne organic chemicals in the whole fish and in different tissues at any time during exposure. The predictive power of the model was tested against experimental data for six chemicals with n-octanol-water partitioning coefficient (log K ow ) values ranging from 2.13–4.29. Model performance was excellent, with a root mean squared error of 0.28 log units. This model has the potential to help identify suitable habitats for restocking under eel management plans. - Highlights: • A PBTK model was developed for European eel (Anguilla anguilla). • Own experimental data and data from the literature were used for parameterization. • The predictive power of the model was excellent, with RMSE of 0.28 log units. • The developed model can be amended with sub-models for dietary and dermal exposure

  20. Calculational experimental examination and ensuring of equipment and pipelines seismic resistance at starting and operating water-cooled and moderated reactor WWER-type NPPs. Final report

    International Nuclear Information System (INIS)

    1999-01-01

    The results of testing of equipment at Bohunice NPP and pipeline systems at Unit 3 of Kozloduy NPP (WWER-440 type reactors) are presented in this Final Report. These results side by side with experimental values of natural frequencies and decrements also include experimental data about vibration modes of tested equipment and pipelines. For the first time the results of new calculational-experimental examination of equipment seismic resistance at Unit 2 of Armenian NPP are presented. At Kozloduy NPP direction's request the planed additional tests of some selected items were put off on 1997. Instead of postponed tests we carried out detailed analysis of our past inspections of numerous equipment seismic resistance at the Unit 5 of Kozloduy NPP. Experimental data with results of additional analysis are presented

  1. Design of a target and moderator at the Los Alamos Spallation Radiation Effects Facility (LASREF) as a neutron source for fusion reactor materials development

    International Nuclear Information System (INIS)

    Ferguson, P.D.; Mueller, G.E.

    1993-01-01

    The LASREF facility is located in the beam stop area at LAMPF. The neutron spectrum is fission-like with the addition of a 3% to 5% component with E > 20 MeV. The present study evaluates the limits on geometry and material selection that will maximize the neutron flux. MCNP and LAHET were used to predict the neutron flux and energy spectrum for a variety of geometries. The problem considers 760 MeV protons incident on tungsten. The resulting neutrons are multiplied in uranium through (n,xn) reactions. Calculations show that a neutron flux greater than 10 19 n/m 2 /s is achievable. The helium to dpa ratio and the transmutation product generation are calculated. These results are compared to expectations for the proposed DEMO fusion reactor and to FFTF

  2. The influence of a spatial displacement of hydrogen on the reactivity and neutron flux density distribution in a ZrH-moderated reactor

    International Nuclear Information System (INIS)

    Doehler, J.; Bartsch, G.

    1975-08-01

    The effect of changes of the hydrogen concentration in uranium zirconium hydride resulting from spatially varying temperatures on the reactivity and neutron flux distribution of the BER-II core (power 2.2 MW) are shown. Furthermore, in general, the influence of the hydrogen concentration on important reactor parameters of a fuel cell of BER-II is calculated and presented. A comparison of the diffusion calculation with spatially constant hydrogen concentrations shows a decrease of the thermal neutron flux density in regions with a low hydrogen content (high temperature) and inversely an increase for high hydrogen concentrations. Furthermore, a change of the effective multiplication factor by 0.6% was found in the case of a spatially varying hydrogen concentration as compared with the calculation for a constant concentration. (orig.) [de

  3. Assessment of competing reaction effect on results of activation analysis with use of water-cooled and water-moderated reactor neutron fields

    International Nuclear Information System (INIS)

    Avsaragov, Kh.B.; Toichkin, A.N.; Lobov, A.N.

    1988-01-01

    Effect of competing threshold reactions on results of neutron activation analysis (NAA) using WWER-440 reactor is investigated. (n,p) and (n,α) fast neutron and 232 Th (n,f), 235 U(n,f), 238 U(n,f) fast and thermal neutron processes are considered as competing ones. Contribution of competing reactions when determining Na, Mn, Sc, Fe, Cu, Y for the core channels of in-core monitoring and ionization chamber ring water protection is experimentally evaluated using a spectrometer with Ge(Li) detector in a set with AI-4096 analyser. Under rigid neutron fields interfering activity increases at the expense of thorium and uranium atom fission. It is stressed that when determining Zr, Mo, Ru, Ba, La, Ce, Nd contribution of fission reaction products can appear to be sufficient

  4. Operational strategies for thermophilic anaerobic digestion of organic fraction of municipal solid waste in continuously stirred tank reactors

    DEFF Research Database (Denmark)

    Angelidaki, Irini; Cui, J.; Chen, X.

    2006-01-01

    Three operational strategies to reduce inhibition due to ammonia during thermophilic anaerobic digestion of source-sorted organic fraction of municipal solid waste (SS-OFMSW) rich in proteins were investigated. Feed was prepared by diluting SS-OFMSW (ratio of 1:4) with tap water or reactor process...... ammonium bicarbonate additions. Dilution of SS-OFMSW with fresh water showed a stable performance with volatile fatty acids of solids (VS). Use of recirculated process water after stripping ammonia showed even better performance with a methane yield...... of 0.43 m(3) kg(-1)VS. Recirculation of process water alone on the other hand, resulted in process inhibition at both TAN levels of 3.5 and 5.5 g-N l(-1). However, after a short period, the process recovered and adapted to the tested TAN levels. Thus, use of recirculated process water after stripping...

  5. Impact of radiolysis and radiolytic corrosion on the release of {sup 13}C and {sup 37}Cl implanted into nuclear graphite: Consequences for the behaviour of {sup 14}C and {sup 36}Cl in gas cooled graphite moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moncoffre, N., E-mail: nathalie.moncoffre@ipnl.in2p3.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Toulhoat, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Silbermann, G. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); and others

    2016-04-15

    Graphite finds widespread use in many areas of nuclear technology based on its excellent moderator and reflector qualities as well as its strength and high temperature stability. Thus, it has been used as moderator or reflector in CO{sub 2} cooled nuclear reactors such as UNGG, MAGNOX, and AGR. However, neutron irradiation of graphite results in the production of {sup 14}C (dose determining radionuclide) and {sup 36}Cl (long lived radionuclide), these radionuclides being a key issue regarding the management of the irradiated waste. Whatever the management option (purification, storage, and geological disposal), a previous assessment of the radioactive inventory and the radionuclide's location and speciation has to be made. During reactor operation, the effects of radiolysis are likely to promote the radionuclide release especially at the gas/graphite interface. Radiolysis of the coolant is mainly initiated through γ irradiation as well as through Compton electrons in the graphite pores. Radiolysis can be simulated in laboratory using γ irradiation or ion irradiation. In this paper, {sup 13}C, {sup 37}Cl and {sup 14}N are implanted into virgin nuclear graphite in order to simulate respectively the presence of {sup 14}C, {sup 36}Cl and nitrogen, a {sup 14}C precursor. Different irradiation experiments were carried out using different irradiation devices on implanted graphite brought into contact with a gas simulating the coolant. The aim was to assess the effects of gas radiolysis and radiolytic corrosion induced by γ or He{sup 2+} irradiation at the gas/graphite interface in order to evaluate their role on the radionuclide release. Our results allow inferring that radiolytic corrosion has clearly promoted the release of {sup 14}C, {sup 36}Cl and {sup 14}N located at the graphite brick/gas interfaces and open pores.

  6. Organic loading rate impact on biohydrogen production and microbial communities at anaerobic fluidized thermophilic bed reactors treating sugarcane stillage.

    Science.gov (United States)

    Santos, Samantha Christine; Rosa, Paula Rúbia Ferreira; Sakamoto, Isabel Kimiko; Varesche, Maria Bernadete Amâncio; Silva, Edson Luiz

    2014-05-01

    This study aimed to evaluate the effect of high organic loading rates (OLR) (60.0-480.00 kg COD m(-3)d(-1)) on biohydrogen production at 55°C, from sugarcane stillage for 15,000 and 20,000 mg CODL(-1), in two anaerobic fluidized bed reactors (AFBR1 and AFBR2). It was obtained, for H2 yield and content, a decreasing trend by increasing the OLR. The maximum H2 yield was observed in AFBR1 (2.23 mmol g COD added(-1)). The volumetric H2 production was proportionally related to the applied hydraulic retention time (HRT) of 6, 4, 2 and 1h and verified in AFBR1 the highest value (1.49 L H2 h(-1)L(-1)). Among the organic acids obtained, there was a predominance of lactic acid (7.5-22.5%) and butyric acid (9.4-23.8%). The microbial population was set with hydrogen-producing fermenters (Megasphaera sp.) and other organisms (Lactobacillus sp.). Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Exergoeconomic assessment and parametric study of a Gas Turbine-Modular Helium Reactor combined with two Organic Rankine Cycles

    International Nuclear Information System (INIS)

    Mohammadkhani, F.; Shokati, N.; Mahmoudi, S.M.S.; Yari, M.; Rosen, M.A.

    2014-01-01

    An exergoeconomic analysis is reported for a combined system with a net electrical output of 299 MW in which waste heat from a Gas Turbine-Modular Helium Reactor (GT-MHR) is utilized by two Organic Rankine Cycles (ORCs). A parametric study is also done to reveal the effects on the exergoeconomic performance of the combined system of such significant parameters as compressor pressure ratio, turbine inlet temperature, temperatures of evaporators, pinch point temperature difference in the evaporators and degree of superheat at the ORC (Organic Rankine Cycle) turbines inlet. Finally the combined cycle performance is optimized from the viewpoint of exergoeconomics. The results show that the precooler, the intercooler and the ORC condensers exhibit the worst exergoeconomic performance. For the overall system, the exergoeconomic factor, the capital cost rate and the exergy destruction cost rate are determined to be 37.95%, 6876 $/h and 11,242 $/h, respectively. Also, it is observed that the unit cost of electricity produced by the GT-MHR turbine increases with increasing GT-MHR turbine inlet temperature but decreases as the other above mentioned parameters increase. - Highlights: • An exergoeconomic analysis is performed for the GT-MHR/ORC (Organic Rankine Cycle) combined cycle. • The effects of decision parameters on the exergoeconomic performance are studied. • The highest exergy destructions occur in the precooler, intercooler and condenser. • Superheating the working fluid at the ORC turbine inlet is not necessary. • Thermodynamic and exergoeconomic optimal conditions differ from each other

  8. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

    A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

  9. The moderating role of job resources in the relationship between job demands and interleukin-6 in an Italian healthcare organization.

    Science.gov (United States)

    Falco, Alessandra; Dal Corso, Laura; Girardi, Damiano; De Carlo, Alessandro; Comar, Manola

    2018-02-01

    In this study we examined the association between job demands (JD), job resources (JR), and serum levels of a possible biomarker of stress, the pro-inflammatory cytokine interleukin-6 (IL-6). According to the buffer hypothesis of the Job Demands-Resources (JD-R) model, we expected that job resources-defined as job autonomy and social support from supervisor-might buffer the relationship between job demands, defined as emotional demands and interpersonal conflict with colleagues, and IL-6. Data from 119 employees in an Italian public healthcare organization (acute care hospital) were analyzed using multiple regression. In predicting IL-6, the interactions between emotional demands and JR and between interpersonal conflict with colleagues and job autonomy (but not social support) were significant, after controlling for the effect of age and gender. The association between JD and IL-6 was stronger for individuals with low levels of JR, so that levels of IL-6 were highest when JD were high and JR were low. Overall, these results are consistent with the buffer hypothesis of the JD-R model and also extend previous research, showing that the exposure to stressful situations at work, measured as high JD and low JR, is associated with higher levels of IL-6 in hospital employees. © 2017 Wiley Periodicals, Inc.

  10. Gradual crossover in molecular organization of stable liquid H2O at moderately high pressure and temperature

    DEFF Research Database (Denmark)

    Koga, Yoshikata; Westh, Peter; Yoshida, Koh

    2014-01-01

    temperature. The extrapolated temperature to zero p seems to be about 70 – 80 °C for points X and 90 – 110 °C for Y. Furthermore, the mid-points of X and Y seem to extrapolate to the triple point of liquid, ice Ih and ice III. Recalling that the zero x Gly extrapolation of point X and Y for binary aqueous...... of the step named point X and its end point Y. In analogy with another third and fourth derivative pair in binary aqueous solutions of glycerol, dα p /dx Gly and d2 α p /dx Gly 2, at 0.1 MPa (α p is the thermal expansivity and x Gly the mole fraction of solute glycerol) in our recent publication [J. Solution...... Chem. 43, 663-674 (2014); DOI:10.1007/s10953-013-0122-7], we argue that there is a gradual crossover in the molecular organization of pure H2O from a low to a high p-regions starting at point X and ending at Y at a fixed T. The crossover takes place gradually spanning for about 100 MPa at a fixed...

  11. Mechanical spectral shift reactor

    International Nuclear Information System (INIS)

    Doshi, P.K.; George, R.A.; Dollard, W.J.

    1982-01-01

    A mechanical spectral shift arrangement for controlling a nuclear reactor includes a plurality of reactor coolant displacer members which are inserted into a reactor core at the beginning of the core life to reduce the volume of reactor coolant-moderator in the core at start-up. However, as the reactivity of the core declines with fuel depletion, selected displacer members are withdrawn from the core at selected time intervals to increase core moderation at a time when fuel reactivity is declining. (author)

  12. Effectiveness of inorganic and organic mulching for soil salinity and sodicity control in a grapevine orchard drip-irrigated with moderately saline waters

    Directory of Open Access Journals (Sweden)

    Ramón Aragüés

    2014-05-01

    Full Text Available Soil mulching is a sensible strategy to reduce evaporation, accelerate crop development, reduce erosion and assist in weed control, but its efficiency for soil salinity control is not as well documented. The benefits of inorganic (plastic and organic (grapevine pruning residues mulching for soil salinity and sodicity control were quantified in a grapevine orchard (cultivars ‘Autumn’ Royal and ‘Crimson’ drip-irrigated with moderately saline waters. Soil samples were taken at the beginning and end of the 2008 and 2009 irrigation seasons in six vines of each cultivar and mulching treatment. Soil saturation extract electrical conductivity (ECe, chloride (Cle and sodium adsorption ratio (SARe values increased in all treatments of both grapevines along the irrigation seasons, but the increases were much lower in the mulched than in the bare soils due to reduced evaporation losses and concomitant decreases in salt evapo-concentration. The absolute salinity and sodicity daily increases in ‘Autumn’ and ‘Crimson’ 2008 and in ‘Crimson’ 2009 were on the average 44% lower in the plastic and 76% lower in the organic mulched soils than in the bare soil. The greater efficiency of the organic than the plastic mulch in ‘Crimson’ 2009 was attributed to the leaching of salts by a precipitation of 104 mm that infiltrated the organic mulch but was intercepted by the plastic mulch. Although further work is needed to substantiate these results, the conclusion is that the plastic mulch and, particularly, the organic mulch were more efficient than the bare soil for soil salinity and sodicity control.

  13. Thermophilic two-phase anaerobic digestion using an innovative fixed-bed reactor for enhanced organic matter removal and bioenergy recovery from sugarcane vinasse

    International Nuclear Information System (INIS)

    Fuess, Lucas Tadeu; Kiyuna, Luma Sayuri Mazine; Ferraz, Antônio Djalma Nunes; Persinoti, Gabriela Felix

    2017-01-01

    Highlights: • An innovative fixed-film anaerobic reactor was applied to sugarcane vinasse. • Stable operation was observed for OLRs as high as 30 kg COD m"−"3 day"−"1. • Propionate buildup did not impact the stability of the structured-bed reactor. • Enhanced bioenergy recovery was estimated from biodigestion with phase separation. • Energy extraction was over 20% higher compared to single-phase systems. - Abstract: This study considered the application of anaerobic digestion (AD) with phase separation combined with the use of an anaerobic structured-bed reactor (ASTBR) as the methanogenic phase for the treatment of sugarcane vinasse, a high-strength wastewater resulting from ethanol production. Two combined thermophilic acidogenic-methanogenic systems formed by one single acidogenic reactor followed by two methanogenic reactors operated in parallel were compared, namely, a conventional UASB reactor and an upflow ASTBR reactor. Increasing organic loading rate (OLR) conditions (15–30 kg COD m"−"3 d"−"1) were applied to the methanogenic reactors. The results highlighted the feasibility of applying the ASTBR to vinasse, indicating a global COD removal higher than 80%. The ASTBR exhibited a stable long-term operation (240 days), even for OLR values as high as 30 kg COD m"−"3 d"−"1. The application of similar conditions to the UASB reactor indicated severe performance losses, leading to the accumulation of acids for every increase in the OLR. An energetic potential of 181.5 MJ for each cubic meter of vinasse was estimated from both hydrogen and methane. The provision of bicarbonate alkalinity proved to be a key factor in obtaining stable performance, offsetting the limitations of relatively low hydraulic retention times (<24 h).

  14. Interactions between multiple organizations responding to a reactor accident in Switzerland

    International Nuclear Information System (INIS)

    Baggenstos, M.; Isaak, H.P.

    1987-01-01

    The poster deals with the problems that emerge from the distribution of responsibilities in the case of a nuclear accident. It is demonstrated how clear and simple procedures and good co-ordination between the onsite and offsite emergency organizations are necessary for the smooth handling of such a situation. A good communications network is particularly essential. The special problem of co-ordination between different organizations performing measurements of the contaminated air, ground and food is also discussed. From the experience gained during the Chernobyl accident, it has been established that only those organizations, which are correctly trained and have a perspective over the complete accident situation, can perform valuable work. At the beginning of the accident for example, there was no organization responsible for taking samples covering the whole of Switzerland. A first quick assessment of the radiological situation could therefore only be made on the basis of measurements performed in the direct vicinity of the laboratories or from single samples which had to be transported over large distances

  15. Gradual crossover in molecular organization of stable liquid H2O at moderately high pressure and temperature

    Directory of Open Access Journals (Sweden)

    Yoshikata Koga

    2014-09-01

    Full Text Available Using the literature raw data of the speed of sound and the specific volume, the isothermal compressibility, κT, a second derivative thermodynamic quantity of G, was evaluated for liquid H2O in the pressure range up to 350 MPa and the temperature to 50 ºC. We then obtained its pressure derivative, dκT/dp, a third derivative numerically without using a fitting function to the κT data. On taking yet another p-derivative at a fixed T graphically without resorting to any fitting function, the resulting d2κT/dp2, a fourth derivative, showed a weak but clear step anomaly, with the onset of the step named point X and its end point Y. In analogy with another third and fourth derivative pair in binary aqueous solutions of glycerol, dαp/dxGly and d2αp/dxGly2, at 0.1 MPa (αp is the thermal expansivity and xGly the mole fraction of solute glycerol in our recent publication [J. Solution Chem. 43, 663-674 (2014; DOI:10.1007/s10953-013-0122-7], we argue that there is a gradual crossover in the molecular organization of pure H2O from a low to a high p-regions starting at point X and ending at Y at a fixed T. The crossover takes place gradually spanning for about 100 MPa at a fixed temperature. The extrapolated temperature to zero p seems to be about 70 – 80 °C for points X and 90 – 110 °C for Y. Furthermore, the mid-points of X and Y seem to extrapolate to the triple point of liquid, ice Ih and ice III. Recalling that the zero xGly extrapolation of point X and Y for binary aqueous glycerol at 0.1 MPa gives about the same T values respectively, we suggest that at zero pressure the region below about 70 °C the hydrogen bond network is bond-percolated, while above about 90 ºC there is no hydrogen bond network. Implication of these findings is discussed.

  16. Electrochemical oxidation of recalcitrant organic compounds in biologically treated municipal solid waste leachate in a flow reactor.

    Science.gov (United States)

    Quan, Xuejun; Cheng, Zhiliang; Chen, Bo; Zhu, Xincai

    2013-10-01

    Biologically-treated municipal solid waste (MSW) leachate still contains many kinds of bio-recalcitrant organic matter. A new plate and frame electrochemical reactor was designed to treat these materials under flow conditions. In the electrochemical oxidation process, NH3 and color could be easily removed by means of electro-generated chlorine/hypochlorite within 20 min. The effects of major process parameters on the removal of organic pollutants were investigated systematically. Under experimental conditions, the optimum operation parameters were current density of 65 mA/cm2, flow velocity of 2.6 cm/sec in electrode gap, and initial chloride ion concentration of 5000 mg/L. The COD in the leachate could be reduced below 100 mg/L after 1 hr of treatment. The kinetics and mechanism of COD removal were investigated by simultaneously monitoring the COD change and chlorine/hypochlorite production. The kinetics of COD removal exhibited a two-stage kinetic model, and the decrease of electro-generated chlorine/hypochlorite production was the major mechanism for the slowing down of the COD removal rate in the second stage. The narrowing of the electrode gap is beneficial for COD removal and energy consumption.

  17. Evaluation of organic coatings to reduce air leakage through cracks in the Pickering NGS 'A' reactor building 1

    International Nuclear Information System (INIS)

    Deans, J.J.; Sato, J.A.; Hampton, J.H.D.; Cullen, R.; Paterson, G.; Chan, P.; Rajagopalan, R.

    1994-01-01

    Pressure tests conducted in 1992 on the Pickering NGS 'A' Reactor Building 1 showed that the containment leakage rate of the building was close to the licensing limit. The leakage was found to be pressure dependent and was attributed to cracks in the concrete dome. A number of solutions were studied by a task group, and the application of an organic coating to the exterior surface of the dome was identified as the most viable solution under the constraints of schedule and cost. In addition to reducing the air leakage rate, the coating material must be flexible to bridge existing moving cracks, it must have excellent adhesion to the concrete substrate to sustain the design pressure of 41.4 kPa(g) during pressure tests, and it must be durable for an exterior application and service conditions. Five candidate organic coating materials were selected for laboratory testing. As a result of the testing, a single-component elastomeric polyurethane coating was selected to be used on the dome. This paper discusses the selection process, laboratory tests and results, and the application of the polyurethane coating system to the exterior concrete dome surface. However, the main emphasis of the paper is on the laboratory evaluation of the five candidate materials. (author). 2 refs., 3 tabs., 1 fig

  18. 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

  19. Characterization of aerosol photooxidation flow reactors: heterogeneous oxidation, secondary organic aerosol formation and cloud condensation nuclei activity measurements

    Directory of Open Access Journals (Sweden)

    A. T. Lambe

    2011-03-01

    Full Text Available Motivated by the need to develop instrumental techniques for characterizing organic aerosol aging, we report on the performance of the Toronto Photo-Oxidation Tube (TPOT and Potential Aerosol Mass (PAM flow tube reactors under a variety of experimental conditions. The PAM system was designed with lower surface-area-to-volume (SA/V ratio to minimize wall effects; the TPOT reactor was designed to study heterogeneous aerosol chemistry where wall loss can be independently measured. The following studies were performed: (1 transmission efficiency measurements for CO2, SO2, and bis(2-ethylhexyl sebacate (BES particles, (2 H2SO4 yield measurements from the oxidation of SO2, (3 residence time distribution (RTD measurements for CO2, SO2, and BES particles, (4 aerosol mass spectra, O/C and H/C ratios, and cloud condensation nuclei (CCN activity measurements of BES particles exposed to OH radicals, and (5 aerosol mass spectra, O/C and H/C ratios, CCN activity, and yield measurements of secondary organic aerosol (SOA generated from gas-phase OH oxidation of m-xylene and α-pinene. OH exposures ranged from (2.0 ± 1.0 × 1010 to (1.8 ± 0.3 × 1012 molec cm−3 s. Where applicable, data from the flow tube reactors are compared with published results from the Caltech smog chamber. The TPOT yielded narrower RTDs. However, its transmission efficiency for SO2 was lower than that for the PAM. Transmission efficiency for BES and H2SO4 particles was size-dependent and was similar for the two flow tube designs. Oxidized BES particles had similar O/C and H/C ratios and CCN activity at OH exposures greater than 1011 molec cm−3 s, but different CCN activity at lower OH exposures. The O/C ratio, H/C ratio, and yield of m-xylene and α-pinene SOA was strongly affected by reactor design and

  20. Renormalization group and instantons in stochastic nonlinear dynamics, from self-organized criticality to thermonuclear reactors

    International Nuclear Information System (INIS)

    Volchenkov, D.

    2009-01-01

    Stochastic counterparts of nonlinear dynamics are studied by means of nonperturbative functional methods developed in the framework of quantum field theory (QFT). In particular, we discuss fully developed turbulence, including leading corrections on possible compressibility of fluids, transport through porous media, theory of waterspouts and tsunami waves, stochastic magnetohydrodynamics, turbulent transport in crossed fields, self-organized criticality, and dynamics of accelerated wrinkled flame fronts advancing in a wide canal. This report would be of interest to the broad auditorium of physicists and applied mathematicians, with a background in nonperturbative QFT methods or nonlinear dynamical systems, having an interest in both methodological developments and interdisciplinary applications. (author)

  1. Renormalization group and instantons in stochastic nonlinear dynamics, from self-organized criticality to thermonuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Volchenkov, D. [Bielefeld Univ., Center of Excellence Cognitive Interaction Technology (CITEC) (Germany)

    2009-03-15

    Stochastic counterparts of nonlinear dynamics are studied by means of nonperturbative functional methods developed in the framework of quantum field theory (QFT). In particular, we discuss fully developed turbulence, including leading corrections on possible compressibility of fluids, transport through porous media, theory of waterspouts and tsunami waves, stochastic magnetohydrodynamics, turbulent transport in crossed fields, self-organized criticality, and dynamics of accelerated wrinkled flame fronts advancing in a wide canal. This report would be of interest to the broad auditorium of physicists and applied mathematicians, with a background in nonperturbative QFT methods or nonlinear dynamical systems, having an interest in both methodological developments and interdisciplinary applications. (author)

  2. Fusion reactor pumped laser

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1988-01-01

    A nuclear pumped laser is described comprising: a toroidal fusion reactor, the reactor generating energetic neutrons; an annular gas cell disposed around the outer periphery of the reactor, the cell including an annular reflecting mirror disposed at the bottom of the cell and an annular output window disposed at the top of the cell; a gas lasing medium disposed within the annular cell for generating output laser radiation; neutron reflector material means disposed around the annular cell for reflecting neutrons incident thereon back into the gas cell; neutron moderator material means disposed between the reactor and the gas cell and between the gas cell and the neutron reflector material for moderating the energy of energetic neutrons from the reactor; converting means for converting energy from the moderated neutrons to energy pumping means for pumping the gas lasing medium; and beam compactor means for receiving output laser radiation from the annular output window and generating a single output laser beam therefrom

  3. How Environmental Uncertainty Moderates the Effect of Relative Advantage and Perceived Credibility on the Adoption of Mobile Health Services by Chinese Organizations in the Big Data Era

    Directory of Open Access Journals (Sweden)

    Xing Chen

    2016-01-01

    Full Text Available Despite the importance of adoption of mobile health services by an organization on the diffusion of mobile technology in the big data era, it has received minimal attention in literature. This study investigates how relative advantage and perceived credibility affect an organization’s adoption of mobile health services, as well as how environmental uncertainty changes the relationship of relative advantage and perceived credibility with adoption. A research model that integrates relative advantage, perceived credibility, environmental uncertainty, and an organization’s intention to use mobile health service is developed. Quantitative data are collected from senior managers and information systems managers in 320 Chinese healthcare organizations. The empirical findings show that while relative advantage and perceived credibility both have positive effects on an organization’s intention to use mobile health services, relative advantage plays a more important role than perceived credibility. Moreover, environmental uncertainty positively moderates the effect of relative advantage on an organization’s adoption of mobile health services. Thus, mobile health services in environments characterized with high levels of uncertainty are more likely to be adopted because of relative advantage than in environments with low levels of uncertainty.

  4. 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.)

  5. The purification of organic reactor coolants; La purification des refrigerants organiques

    Energy Technology Data Exchange (ETDEWEB)

    Hannaert, H; Lopes Cardozo, R [CCR EURATOM, Ispra, Varese (Italy)

    1967-01-01

    Among the main impurities present in irradiated and virgin coolants we have been particularly interested in chlorine, water, iron, oxygen and heavy elements. Our studies have been directed along two basic lines, namely: (1) the elimination of inorganic impurities, and (2) the elimination of organic impurities. The purpose of the studies on the elimination of inorganic impurities is to obtain a 'clean' coolant (virgin or used), this cleanness being marked by a low tendency to form deposits on the fuel element cladding (fouling corrosion). Careful attention has been paid to the problems of chlorine, water, iron and coated particles. Particular interest has been attached to research on deoxygenation of the organic liquid by catalytic hydrocracking, the oxygen (which comes from the dissociation of the water and the gases contained in the liquid) favouring polymerization and the formation of particles which are likely to be deposited on the hot walls. As regards the elimination of degradation products, many studies have been carried out with a view to permitting the maximum recycling of decomposed hydrocarbides and thus reducing the cost of make-up, the recycling rate being a function of the mean molecular weight and the viscosity of the recycled fractions. The technique based on solvent extraction has led to very satisfactory results. An extension of this technique, using counter-flow extraction, appears to be even more promising. (author)

  6. 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

  7. Artificial neural network with self-organizing mapping for reactor stability monitoring

    International Nuclear Information System (INIS)

    Okumura, Motofumi; Tsuji, Masashi; Shimazu, Yoichiro; Narabayashi, Tadashi

    2008-01-01

    In BWR stability monitoring damping ratio has been used as a stability index. A method for estimating the damping ratio by applying Principal Component Analysis (PCA) to neutron detector signals measured with local power range monitors (LPRMs) had been developed; In this method, measured fluctuating signal is decomposed into some independent components and the signal component directly related to stability is extracted among them to determine the damping ratio. For online monitoring, it is necessary to select stability related signal component efficiently. The self-organizing map (SOM) is one of the artificial neural networks and has the characteristics such that online learning is possible without supervised learning within a relatively short time. In the present study, the SOM was applied to extract the relevant signal component more quickly and more accurately, and the availability was confirmed through the feasibility study. (author)

  8. Person-organization fit and turnover intention: exploring the mediating effect of work engagement and the moderating effect of demand-ability fit.

    Science.gov (United States)

    Peng, Jui-Chen; Lee, Yin-Ling; Tseng, Mei-Man

    2014-03-01

    Person-organization (P-O) fit is an important influencing factor on the intentions and attitudes of hospital nurses. The authors used a motivation-mechanism approach to conceptualize work engagement as a mediator and demand-ability (D-A) fit as a moderator to elicit the role of P-O fit in the turnover intention of nurses. This article explores whether the work engagement of nurses mediates the relationship between P-O fit and turnover intention and examines whether D-A fit moderates this relationship. The sample comprised 349 nurses working for two regional hospitals in Yilan County, Taiwan. Linear regression modeling analysis was conducted to test the proposed hypotheses. Results indicate that P-O fit has a negative effect on participant turnover intention. In addition, the work engagement of participants was found to mediate the impact of P-O fit on turnover intention. A new significant interactive relationship was discovered such that high D-A fit strengthened the negative relationship between P-O fit and turnover intention. The work engagement of professional nurses has attracted increasing attention in the literature on fit, particularly with regard to the linkage between P-O and fit-turnover intention. This study enhances the understanding of the function of P-O fit by considering perceived D-A fit. Nurse turnover is the main reason for the current shortage of nurses in Taiwan. Therefore, if the cognitive values of nurses and the organizational culture fit with hospital value systems, common values may facilitate a higher degree of nurse work engagement and, in turn, decrease turnover intention. In addition, recruiting employees with high D-A fit may help hospitals enhance the negative relationship between P-O fit and nurse turnover intention.

  9. Sludge granulation in an UASB-moving bed biofilm hybrid reactor for efficient organic matter removal and nitrogen removal in biofilm reactor.

    Science.gov (United States)

    Chatterjee, Pritha; Ghangrekar, M M; Rao, Surampalli

    2018-02-01

    A hybrid upflow anaerobic sludge blanket (UASB)-moving bed biofilm (MBB) and rope bed biofilm (RBB) reactor was designed for treatment of sewage. Possibility of enhancing granulation in an UASB reactor using moving media to improve sludge retention was explored while treating low-strength wastewater. The presence of moving media in the top portion of the UASB reactor allowed a high solid retention time even at very short hydraulic retention times and helped in maintaining selection pressure in the sludge bed to promote formation of different sized sludge granules with an average settling velocity of 67 m/h. These granules were also found to contain plenty of extracellular polymeric substance (EPS) such as 58 mg of polysaccharides (PS) per gram of volatile suspended solids (VSS) and protein (PN) content of 37 mg/g VSS. Enriched sludge of nitrogen-removing bacteria forming a porous biofilm on the media in RBB was also observed in a concentration of around 894 g/m 2 . The nitrogen removing sludge also had a high EPS content of around 22 mg PS/g VSS and 28 mg PN/g VSS. This hybrid UASB-MBB-RBB reactor with enhanced anaerobic granular sludge treating both carbonaceous and nitrogenous matter may be a sustainable solution for decentralized sewage treatment.

  10. Moderate Bravery

    DEFF Research Database (Denmark)

    Majgaard, Klaus

    2016-01-01

    Purpose: The ability to act in a purposeful and effective way amid institutional tensions and paradoxes is, right now, a highly prized quality in public leadership. The purpose of this chapter is to qualify moderately brave acts as a learning format that combines the analytical and performative...

  11. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  12. EDO, Doses to Man and Organs from Reactor Operation Noble Gas and Liquid Waste Release

    International Nuclear Information System (INIS)

    Rodenas Diago, Jose; Serradell Garcia, Vicente

    1983-01-01

    1 - Description of problem or function: EDO evaluates individual and collective doses to man from atmospheric releases of noble gases and other gaseous effluents. 2 - Method of solution: The dose calculations are carried out by following the guide- lines of USNRC Regulatory Guide 1.109. Radiation exposure for maximum individuals and population are estimated within 30 km from the nuclear plant. This area is divided into 160 circular trapezoids, to which computations are referred. Four age groups, seven organs for internal dose and two for external dose have been considered. Dose calculations are done through 14 pathways, 7 for liquid effluents, one for noble gases, and 6 for the rest of gaseous effluents. 3 - Restrictions on the complexity of the problem: The following are the maximum dimension sizes preset in the code: 73 radionuclides (other than noble gases); 15 noble gases; 160 circular trapezoids; 31 chemical elements; 4 types of aquatic foods; 15 points of exposure for shorelines; 15 trapezoids influenced by each point; 4 terrestrial food pathways; 100 centres of population. Some of these limits can be varied

  13. EDO, Doses to Man and Organs from Reactor Operation Noble Gas and Liquid Waste Release

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas Diago, Jose; Serradell Garcia, Vicente [Departamento de Ingenieria Nuclear, Escuela Tecnica Superior de Ingenieros Industriales, Universidad Politecnica, Camino de Vera 2/n Apartado 2012, Valencia (Spain)

    1983-10-18

    1 - Description of problem or function: EDO evaluates individual and collective doses to man from atmospheric releases of noble gases and other gaseous effluents. 2 - Method of solution: The dose calculations are carried out by following the guide- lines of USNRC Regulatory Guide 1.109. Radiation exposure for maximum individuals and population are estimated within 30 km from the nuclear plant. This area is divided into 160 circular trapezoids, to which computations are referred. Four age groups, seven organs for internal dose and two for external dose have been considered. Dose calculations are done through 14 pathways, 7 for liquid effluents, one for noble gases, and 6 for the rest of gaseous effluents. 3 - Restrictions on the complexity of the problem: The following are the maximum dimension sizes preset in the code: 73 radionuclides (other than noble gases); 15 noble gases; 160 circular trapezoids; 31 chemical elements; 4 types of aquatic foods; 15 points of exposure for shorelines; 15 trapezoids influenced by each point; 4 terrestrial food pathways; 100 centres of population. Some of these limits can be varied.

  14. The Dragon reactor experiment

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

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

  15. The stand prototype of minimum power NRE reactor

    International Nuclear Information System (INIS)

    Belogurov, A.I.; Grigorenko, L.N.; Mamontov, Yu.I.; Rachuk, V.S.; Stukalov, A.I.; Konyukhov, G.V.

    1995-01-01

    For ensuring of full-scale development of nuclear rocket engine (NRE) reactor was created stand prototype (reactor IRGIT?) The main differences of its are as follows: 1) Fasteners of technologies channels contents fuel assemblies in bottom are worked out the split. It is provides possibility a distance channels change without disassembly of reactor stand prototype from stand; 2) Cooling of the vessels, the moderator, the reflector and the barrel actuate is carried out by hydrogen; 3) The lower bottom modified for organization the hydrogen efflux in the form a reactor jet; 4) Radiation defence is introduced as part of stand prototype for ensuring of serviceability of stand accessories and tests routine service; 5) Each technology channels is provided of critical nozzle; 6) Control, regulation and defence of reactor has being carried out on stand system

  16. The effects of mediator and granular activated carbon addition on degradation of trace organic contaminants by an enzymatic membrane reactor.

    Science.gov (United States)

    Nguyen, Luong N; Hai, Faisal I; Price, William E; Leusch, Frederic D L; Roddick, Felicity; Ngo, Hao H; Guo, Wenshan; Magram, Saleh F; Nghiem, Long D

    2014-09-01

    The removal of four recalcitrant trace organic contaminants (TrOCs), namely carbamazepine, diclofenac, sulfamethoxazole and atrazine by laccase in an enzymatic membrane reactor (EMR) was studied. Laccases are not effective for degrading non-phenolic compounds; nevertheless, 22-55% removal of these four TrOCs was achieved by the laccase EMR. Addition of the redox-mediator syringaldehyde (SA) to the EMR resulted in a notable dose-dependent improvement (15-45%) of TrOC removal affected by inherent TrOC properties and loading rates. However, SA addition resulted in a concomitant increase in the toxicity of the treated effluent. A further 14-25% improvement in aqueous phase removal of the TrOCs was consistently observed following a one-off dosing of 3g/L granular activated carbon (GAC). Mass balance analysis reveals that this improvement was not due solely to adsorption but also enhanced biodegradation. GAC addition also reduced membrane fouling and the SA-induced toxicity of the effluent. Copyright © 2014 Elsevier Ltd. All rights reserved.

  17. Laboratory Experiments and Modeling for Interpreting Field Studies of Secondary Organic Aerosol Formation Using an Oxidation Flow Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Jose-Luis [Univ. of Colorado, Boulder, CO (United States)

    2016-02-01

    This grant was originally funded for deployment of a suite of aerosol instrumentation by our group in collaboration with other research groups and DOE/ARM to the Ganges Valley in India (GVAX) to study aerosols sources and processing. Much of the first year of this grant was focused on preparations for GVAX. That campaign was cancelled due to political reasons and with the consultation with our program manager, the research of this grant was refocused to study the applications of oxidation flow reactors (OFRs) for investigating secondary organic aerosol (SOA) formation and organic aerosol (OA) processing in the field and laboratory through a series of laboratory and modeling studies. We developed a gas-phase photochemical model of an OFR which was used to 1) explore the sensitivities of key output variables (e.g., OH exposure, O3, HO2/OH) to controlling factors (e.g., water vapor, external reactivity, UV irradiation), 2) develop simplified OH exposure estimation equations, 3) investigate under what conditions non-OH chemistry may be important, and 4) help guide design of future experiments to avoid conditions with undesired chemistry for a wide range of conditions applicable to the ambient, laboratory, and source studies. Uncertainties in the model were quantified and modeled OH exposure was compared to tracer decay measurements of OH exposure in the lab and field. Laboratory studies using OFRs were conducted to explore aerosol yields and composition from anthropogenic and biogenic VOC as well as crude oil evaporates. Various aspects of the modeling and laboratory results and tools were applied to interpretation of ambient and source measurements using OFR. Additionally, novel measurement methods were used to study gas/particle partitioning. The research conducted was highly successful and details of the key results are summarized in this report through narrative text, figures, and a complete list of publications acknowledging this grant.

  18. In-core LOCA (PTR) analysis with poisoned moderator

    International Nuclear Information System (INIS)

    Kim, S. R.; Kim, B. G.; Kim, T. M.; Choi, J. H.; Kim, Yun Ho; Choi, Hoon

    2005-01-01

    CANDU reactors have been analyzed and evaluated for the postulated in-core LOCA while the reactor is operating normally with low moderator poison concentration. However, when the reactor is operating with relatively large amounts of boron and/or gadolinium poisons in the moderator, the assessment for fuel integrity was required for pressure tube rupture (PTR) accident. The methodology of in-core LOCA analysis with poisoned moderator is developed to determine the effective trip parameters, evaluate the fuel integrity, and establish the standard reactor start-up model for CANDU reactor recently. The developed methodology and results are presented

  19. Cold moderators at ORNL

    International Nuclear Information System (INIS)

    Lucas, A. T.

    1997-09-01

    The Advanced Neutron Source (ANS) cold moderators were not an 'Oak Ridge first', but would have been the largest both physically and in terms of cold neutron flux. Two cold moderators were planned each 410 mm in diameter and containing about 30L of liquid deuterium. They were to be completely independent of each other. A modular system design was used to provide greater reliability and serviceability. When the ANS was terminated, upgrading of the resident High Flux Isotope Reactor (HFIR) was examined and an initial study was made into the feasibility of adding a cold source. Because the ANS design was modular, it was possible to use many identical design features. Sub-cooled liquid at 4 bar abs was initially chosen for the HFIR design concept, but this was subsequently changed to 15 bar abs to operate above the critical pressure. As in the ANS, the hydrogen will operate at a constant pressure throughout the temperature range and a completely closed loop with secondary containment was adopted. The heat load of 2 kW made the heat flux comparable with that of the ANS. Subsequent studies into the construction of cryogenic moderators for the proposed new Synchrotron Neutron source indicated that again many of the same design concepts could be used. By connecting the two cold sources together in series, the total heat load of 2 kW is very close to that of the HFIR allowing a very similar supercritical hydrogen system to be configured. The two hydrogen moderators of the SNS provide a comparable heat load to the HFIR moderator. It is subsequently planned to connect the two in series and operate from a single cold loop system, once again using supercritical hydrogen. The spallation source also provided an opportunity to re-examine a cold pellet solid methane moderator operating at 20K.

  20. 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.

  1. Interaction of CREDO [Centralized Reliability Data Organization] with the EBR-II [Experimental Breeder Reactor II] PRA [probabilistic risk assessment] development

    International Nuclear Information System (INIS)

    Smith, M.S.; Ragland, W.A.

    1989-01-01

    The National Academy of Sciences review of US Department of Energy (DOE) class 1 reactors recommended that the Experimental Breeder Reactor II (EBR-II), operated by Argonne National Laboratory (ANL), develop a level 1 probabilistic risk assessment (PRA) and make provisions for level 2 and level 3 PRAs based on the results of the level 1 PRA. The PRA analysis group at ANL will utilize the Centralized Reliability Data Organization (CREDO) at Oak Ridge National Laboratory to support the PRA data needs. CREDO contains many years of empirical liquid-metal reactor component data from EBR-II. CREDO is a mutual data- and cost-sharing system sponsored by DOE and the Power Reactor and Nuclear Fuels Development Corporation of Japan. CREDO is a component based data system; data are collected on components that are liquid-metal specific, associated with a liquid-metal environment, contained in systems that interface with liquid-metal environments, or are safety related for use in reliability/availability/maintainability (RAM) analyses of advanced reactors. The links between the EBR-II PRA development effort and the CREDO data collection at EBR-II extend beyond the sharing of data. The PRA provides a measure of the relative contribution to risk of the various components. This information can be used to prioritize future CREDO data collection activities at EBR-II and other sites

  2. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

    2000-01-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  3. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor; Etude des consequences de la rupture d'un tube de force dans la cuve d'un reacteur modere a l'eau lourde et refroidi au gaz

    Energy Technology Data Exchange (ETDEWEB)

    Hareux, F; Roche, R; Vrillon, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [French] L'eclatement d'un tube de force dans la cuve d'un reacteur de puissance modere a l'eau lourde et refroidi par un gaz sous pression est un accident qui a ete etudie experimentalement a propos d'EL-4. Un premier essai a l'echelle 1 ayant montre que l'eclatement d'un tube ne provoque pas celui des tubes voisins, des essais relatifs a la tenue de la cuve ont ete effectues sur maquettes a echelle tres reduite (l/lO). Il a ete trouve que la cuve peut supporter plusieurs eclatements de tubes sans deformations notables. La pression transitoire dans la cuve a une allure oscillatoire amortie dont le maximum (pression de pic) a fait l'objet d'une etude experimentale detaillee. Il apparait que les parametres essentiels influant sur cette pression sont: la pression du gaz contenu dans le tube de force, le volume du gaz qui participe a l'eclatement, la flexibilite de la cuve, la masse d'air empoisonnee dans la cuve, la nature du gaz explosant. Une methode generale d'estimation des pics de pression dans

  4. Reduction of the hydraulic retention time at constant high organic loading rate to reach the microbial limits of anaerobic digestion in various reactor systems.

    Science.gov (United States)

    Ziganshin, Ayrat M; Schmidt, Thomas; Lv, Zuopeng; Liebetrau, Jan; Richnow, Hans Hermann; Kleinsteuber, Sabine; Nikolausz, Marcell

    2016-10-01

    The effects of hydraulic retention time (HRT) reduction at constant high organic loading rate on the activity of hydrogen-producing bacteria and methanogens were investigated in reactors digesting thin stillage. Stable isotope fingerprinting was additionally applied to assess methanogenic pathways. Based on hydA gene transcripts, Clostridiales was the most active hydrogen-producing order in continuous stirred tank reactor (CSTR), fixed-bed reactor (FBR) and anaerobic sequencing batch reactor (ASBR), but shorter HRT stimulated the activity of Spirochaetales. Further decreasing HRT diminished Spirochaetales activity in systems with biomass retention. Based on mcrA gene transcripts, Methanoculleus and Methanosarcina were the predominantly active in CSTR and ASBR, whereas Methanosaeta and Methanospirillum activity was more significant in stably performing FBR. Isotope values indicated the predominance of aceticlastic pathway in FBR. Interestingly, an increased activity of Methanosaeta was observed during shortening HRT in CSTR and ASBR despite high organic acids concentrations, what was supported by stable isotope data. Copyright © 2016 Elsevier Ltd. All rights reserved.

  5. Validation of moderator-level reactivity coefficient using station data

    Energy Technology Data Exchange (ETDEWEB)

    Younis, M.; Martchouk, I., E-mail: mohamed.younis@amecfw.com, E-mail: iouri.martchouk@amecfw.com [Amec Foster Wheeler, Toronto, ON (Canada); Buchan, P.D., E-mail: david.buchan@opg.com [Ontario Power Generation, Pickering, ON (Canada)

    2015-07-01

    The reactivity effect due to variations in the moderator level has been recognized as a reactor physics phenomenon of importance during normal operation and accident analysis. The moderator-level reactivity coefficient is an important parameter in safety analysis of CANDU reactors, e.g., during Loss of Moderator Heat Sink as well as in the simulation of Reactor Regulating System action in CANDU reactors that use moderator level for reactivity control. This paper presents the results of the validation exercise of the reactor-physics toolset using the measurements performed in Pickering Unit 4 in 2003. The capability of the code suite of predicting moderator-level reactivity effect was tested by comparing measured and predicted reactor-physics parameters. (author)

  6. Transmutation of Tc-99 in fission reactors

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Li, J.M.

    1994-12-01

    Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

  7. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

    Purpose: To remove the heat of reactor core using a great amount of moderators at the periphery of the reactor core as coolants. Constitution: Heat of a reactor core is removed by disposing a spontaneous recycling cooling device for cooling moderators in a moderator tank, without using additional power driven equipments. That is, a spontaneous recycling cooling device for cooling the moderators in the moderator tank is disposed. Further, the gap between the inner wall of a pressure tube guide pipe disposed through the vertical direction of a moderator tank and the outer wall of a pressure tube inserted through the guide pipe is made smaller than the rupture distortion caused by the thermal expansion upon overheating of the pressure tube and greater than the minimum gap required for heat shiels between the pressure tube and the pressure tube guide pipe during usual operation. In this way, even if such an accident as can not using a coolant cooling device comprising power driven equipment should occur in the pressure tube type reactor, the rise in the temperature of the reactor core can be retarded to obtain a margin with time. (Kamimura, M.)

  8. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  9. Development and Application of an Oxidation Flow Reactor to Study Secondary Organic Aerosol Formation from Ambient Air

    Science.gov (United States)

    Palm, Brett Brian

    Secondary organic aerosols (SOA) in the atmosphere play an important role in air quality, human health, and climate. However, the sources, formation pathways, and fate of SOA are poorly constrained. In this dissertation, I present development and application of the oxidation flow reactor (OFR) technique for studying SOA formation from OH, O3, and NO3 oxidation of ambient air. With a several-minute residence time and a portable design with no inlet, OFRs are particularly well-suited for this purpose. I first introduce the OFR concept, and discuss several advances I have made in performing and interpreting OFR experiments. This includes estimating oxidant exposures, modeling the fate of low-volatility gases in the OFR (wall loss, condensation, and oxidation), and comparing SOA yields of single precursors in the OFR with yields measured in environmental chambers. When these experimental details are carefully considered, SOA formation in an OFR can be more reliably compared with ambient SOA formation processes. I then present an overview of what OFR measurements have taught us about SOA formation in the atmosphere. I provide a comparison of SOA formation from OH, O3, and NO3 oxidation of ambient air in a wide variety of environments, from rural forests to urban air. In a rural forest, the SOA formation correlated with biogenic precursors (e.g., monoterpenes). In urban air, it correlated instead with reactive anthropogenic tracers (e.g., trimethylbenzene). In mixed-source regions, the SOA formation did not correlate well with any single precursor, but could be predicted by multilinear regression from several precursors. Despite these correlations, the concentrations of speciated ambient VOCs could only explain approximately 10-50% of the total SOA formed from OH oxidation. In contrast, ambient VOCs could explain all of the SOA formation observed from O3 and NO3 oxidation. Evidence suggests that lower-volatility gases (semivolatile and intermediate-volatility organic

  10. Pressure tube reactor

    International Nuclear Information System (INIS)

    Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

    1993-01-01

    In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

  11. Research reactors and materials testing

    International Nuclear Information System (INIS)

    Vidal, H.

    1986-01-01

    Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

  12. Metal burning in graphite-moderated reactors

    International Nuclear Information System (INIS)

    Wichner, R.P.; Ball, S.J.; Daw, C.S.; Thomas, J.F.

    1997-01-01

    Pinto beans, sweet corn, and zucchini squash (Cucurbita pepo var. black beauty) were grown in a randomized complete-block field/pot experiment at a site that contained the highest observed levels of surface gross gamma radioactivity within Los Alamos Canyon (LAC) at Los Alamos National Laboratory. Soils as well as washed edible and nonedible crop tissues were analyzed for various radionuclides and heavy metals. Most radionuclides, with the exception of 3 H and tot U, in soil from LAC were detected in significantly higher concentrations (p -1 . This dose was below the International Commission on Radiological Protection permissible dose limit (PDL) of 100 mrem y -1 from all pathways; however, the addition of other internal and external exposure route factors may increase the overall dose over the PDL. Also, the risk of an excess cancer fatality, based on 74 mrem y -1 , was 3.7 x 10 -5 (37 in a million), which is above the Environmental Protection Agency's (acceptable) guideline of one in a million. 25 refs

  13. Effect of the temperature and of the organic load in two-stage up flow anaerobic sludge blanket reactors treating of swine wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Bichuette, Alexandre Abud; Duda, Rose Maria; Oliveira, Roberto Alves de [Universidade Estadual Paulista (UNESP), Jaboticabal, SP (Brazil). Dept. de Engenharia Rural], E-mail: oliveira@fcav.unesp.br

    2008-07-01

    In this work the acting of two-stage up flow anaerobic sludge blanket reactors (UASB) was evaluated, installed in series, in pilot scale (volumes of 908 L and 350 L, respectively) in the treatment swine wastewater, with concentrations of total solids suspended (TSS) around 10000 mg L{sup -1}. The organic loading rates (OLR) applied in first UASB were of 5,2 and of 8,6 g total COD (Ld){sup -1}. The medium efficiencies of removal of the chemical demand of total oxygen (total COD), TSS and TKN were higher than 89; 80 and 55%, respectively, for the system of anaerobic treatment composed by the reactors UASB in two apprenticeships. The rate of volumetric methane production in the system of anaerobic treatment with the reactors UASB were 0,08 and 0,16 m{sup 3}CH{sub 4} (m{sup 3} CH{sub 4} reactor d){sup -1}. The number of total coliforms was reduced to 2,6x10{sup 4} NMP/100 mL. (author)

  14. Graphite moderated {sup 252}Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Sajo B, L.; Barros, H.; Greaves, E. D. [Universidad Simon Bolivar, Nuclear Physics Laboratory, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of); Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a {sup 252}Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the {sup 252}Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

  15. Reactors of different types in the world nuclear power

    International Nuclear Information System (INIS)

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

  16. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators 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 objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  17. 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

  18. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  19. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

    1987-01-01

    Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

  20. Determination of reactor parameters in a simulated RA reactor lattice by measuring the reactivity level of heavy water in D{sub 2}O moderated RB reactor; Odredjivanje reaktorskih parametara u simuliranoj resetki reaktora RA merenjem reaktivnosti nivoa teske vode u D{sub 2}O moderiranom reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Takac, S M; Markovic, H D; Dimitrijevic, Z B [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1974-07-01

    Direct method for experimental determination of the neutron age {tau} in the reactor lattice is not developed. Fermi theory for determining {tau} by measuring the reactivity level of the heavy water can be applied for a limited number of reactor lattices. An attempt was made to apply this approach for a complex reactor core with side and upper reflector. As expected the obtained results were not satisfactory, and {tau} was determined by Dessauer formula which gives more realistic estimation of thermalization in the reactor cell. major discrepancies are resulting from the fact that the lower reflector was neglected. But it is possible to to determine {tau} for reactor core with reflectors by additional measurements of axial distribution and other experimental data for {tau}. This is quite tedious numerical procedure. Obtained experimental data for a number of reactor parameters are compared to the initial data of the RA reactor core showing very good agreement. [Serbo-Croat] Ekspeimentalna metoda za direktno odredjivanje starosti neutrona {tau}, u resetki reaktora do danas nije razvijena. Primenjena Fermijeova teorija za odredjivanje {tau} preko merenja reaktivnosti nivoa teske vode, ocigledno se moze koristiti samo na ogranicen spektar reaktorskih resetki. U ovom radu pokusano je da se vidi primenljivost iste u slozenom - bocno i odozdo - reflektovanom jezgru reaktora. Naravno dobijeni rezultati nisu zadovoljili, sto se i moglo ocekivati, pa je {tau} odredjen preko Dessauer-ove formule, sto daje daleko blizu sliku stvarnog stanja procesa termalizacije u celiji reaktora. Ocigledno vece neslaganje primenjene teorije dolazi od zanemarivanja donjeg reflektora u jezgru reaktora. Medjutim, dopnskim merenjem aksijalne raspodele i na osnovu ostalih eksperimentalnih podataka za {tau}, moguce je odrediti eksperimentalno reflektorski koeficijent za visestruko reflektovano jezgro reaktora, sto numericki predstavlja mukotrpan i dugotrajan rad na racunaru. Dobijeni

  1. Russian RBMK reactor design information

    International Nuclear Information System (INIS)

    1993-11-01

    This document concerns the systems, design, and operations of the graphite-moderated, boiling, water-cooled, channel-type (RBMK) reactors located in the former Soviet Union (FSU). The Russian Academy of Sciences Nuclear Safety Institute (NSI) in Moscow, Russia, researched specific technical questions that were formulated by the Pacific Northwest Laboratory (PNL) and provided detailed technical answers to those questions. The Russian response was prepared in English by NSI in a question-and-answer format. This report presents the results of that technical exchange in the context they were received from the NSI organization. Pacific Northwest Laboratory is generating this document to support the US Department of Energy (DOE) community in responding to requests from FSU states, which are seeking Western technological and financial assistance to improve the safety systems of the Russian-designed reactors. This report expands upon information that was previously available to the United States through bilateral information exchanges, international nuclear society meetings, International Atomic Energy Agency (IAEA) reactor safety programs, and Research and Development Institute of Power Engineering (RDIPE) reports. The response to the PNL questions have not been edited or reviewed for technical consistency or accuracy by PNL staff or other US organizations, but are provided for use by the DOE community in the form they were received

  2. Nuclear reactor shield including magnesium oxide

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1981-01-01

    An improvement is described for nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux. The reactor shielding includes means providing structural support, neutron moderator material, neutron absorber material and other components, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron

  3. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  4. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

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

    International Nuclear Information System (INIS)

    Sartori, Enrico; Blair Briggs, J.

    2008-01-01

    experimental series that were performed at 17 different reactor facilities. The Handbook is organized in a manner that allows easy inclusion of additional evaluations, as they become available. Additional evaluations are in progress and will be added to the handbook periodically. Content: FUND - Fundamental; GCR - Gas Cooled (Thermal) Reactor; HWR - Heavy Water Moderated Reactor; LMFR - Liquid Metal Fast Reactor; LWR - Light Water Moderated Reactor; PWR - Pressurized Water Reactor; VVER - VVER Reactor; Evaluations published as drafts 2 - Related Information: International Criticality Safety Benchmark Evaluation Project (ICSBEP); IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments; IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan ; IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database ; IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility; IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation ; IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility ; IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation ; IRPHE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents; IRPHE-ARCH-01, Archive of HTR Primary Documents ; IRPHE/AVR, AVR High Temperature Reactor Experience, Archival Documentation ; IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters; IRPhE/BERENICE, effective delayed neutron fraction measurements ; IRPhE-TAPIRO-ARCHIVE, fast neutron source reactor primary documents, reactor physics experiments. The International Handbook of Evaluated Reactor Physics Benchmark Experiments was prepared by a working party comprised of experienced reactor physics personnel from Belgium, Brazil, Canada, P.R. of China, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Switzerland, United Kingdom, and the United States of America. The IRPhEP Handbook is available to authorised requesters from the

  6. Chromophoric Dissolved Organic Matter and Dissolved Organic Carbon from Sea-Viewing Wide Field-of-View Sensor (SeaWiFS, Moderate Resolution Imaging Spectroradiometer (MODIS and MERIS Sensors: Case Study for the Northern Gulf of Mexico

    Directory of Open Access Journals (Sweden)

    Blake A. Schaeffer

    2013-03-01

    Full Text Available Empirical band ratio algorithms for the estimation of colored dissolved organic matter (CDOM and dissolved organic carbon (DOC for Sea-viewing Wide Field-of-view Sensor (SeaWiFS, Moderate Resolution Imaging Spectroradiometer (MODIS and MERIS ocean color sensors were assessed and developed for the northern Gulf of Mexico. Match-ups between in situ measurements of CDOM absorption coefficients at 412 nm (aCDOM(412 with that derived from SeaWiFS were examined using two previously reported reflectance band ratio algorithms. Results indicate better performance using the Rrs(510/Rrs(555 (Bias = −0.045; RMSE = 0.23; SI = 0.49, and R2 = 0.66 than the Rrs(490/Rrs(555 reflectance band ratio algorithm. Further, a comparison of aCDOM(412 retrievals using the Rrs(488/Rrs(555 for MODIS and Rrs(510/Rrs(560 for MERIS reflectance band ratios revealed better CDOM retrievals with MERIS data. Since DOC cannot be measured directly by remote sensors, CDOM as the colored component of DOC is utilized as a proxy to estimate DOC remotely. A seasonal relationship between CDOM and DOC was established for the summer and spring-winter with high correlation for both periods (R2~0.9. Seasonal band ratio empirical algorithms to estimate DOC were thus developed using the relationships between CDOM-Rrs and seasonal CDOM-DOC for SeaWiFS, MODIS and MERIS. Results of match-up comparisons revealed DOC estimates by both MODIS and MERIS to be relatively more accurate during summer time, while both of them underestimated DOC during spring-winter time. A better DOC estimate from MERIS in comparison to MODIS in spring-winter could be attributed to its similarity with the SeaWiFS band ratio CDOM algorithm.

  7. Long-term stability of thermophilic co-digestion submerged anaerobic membrane reactor encountering high organic loading rate, persistent propionate and detectable hydrogen in biogas.

    Science.gov (United States)

    Qiao, Wei; Takayanagi, Kazuyuki; Niu, Qigui; Shofie, Mohammad; Li, Yu You

    2013-12-01

    The performance of thermophilic anaerobic co-digestion of coffee grounds and sludge using membrane reactor was i